WorldWideScience

Sample records for palier n4-1450 mwe

  1. Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe Indian PHWR: design perspective

    International Nuclear Information System (INIS)

    Sonavani, Manojkumar; Kelkar, M.G.; Singhvi, P.K.; Roy, S.; Ingle, V.J.

    2013-01-01

    The Flux Mapping System (FMS) of 700 MWe PHWR computes a detailed flux/power distribution of the reactor core using modal synthesis method and is also generate setback on different parameters by monitoring thermal neutron flux at more than 100 points inside the reactor core. These types of setbacks are introduced first time in Indian PHWRs. The paper brings out the Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe and the overall design philosophy. The paper emphasizes on comparisons between 540 MWe and 700 MWe design, considerations for architectural design and setbacks for 700 MWe. (author)

  2. Improved 1500 MWe Arabelle begins operation

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    Two of the first 1500 MWe steam turbine-generator sets, described as the largest in the world, are undergoing commissioning at the Chooz B PWR nuclear power station in France. A number of design improvements have been made over the previous generation of 1350 MWe turbines, a process which will continue. (Author)

  3. Experimental contribution to the study of gas bearings; Contribution experimentale a l'etude des paliers a gaz

    Energy Technology Data Exchange (ETDEWEB)

    Gobert, G

    1962-07-01

    Developments for gas-lubricated bearings are discussed. The study of how the fluid film behaves compared to what one can expect is discussed. Various devices are described which have allowed us to go up to peripheral velocity greater than 400 m/s, leading to rotational speed of several thousand revolutions per second. This is obtained by using an automatic setting for the distance between fixed and rotating walls. (author) [French] Le present travail debute par un apercu de l'etat actuel des connaissances sur le processus de la lubrification, notamment sur les paliers a lubrification par gaz. Apres une analyse effectuee au moyen des recentes theories de R. Comolet sur le film fluide entre disques paralleles, nous decrivons des machines dont les charges axiales et radiales specialement etudiees ont permis, au moyen d'un reglage automatique des jeux entre parois fixes et parois tournantes, d'atteindre des vitesses peripheriques superieures a 400 m/s correspondant a des vitesses de plusieurs milliers de tours par seconde. (auteur)

  4. Experimental contribution to the study of gas bearings; Contribution experimentale a l'etude des paliers a gaz

    Energy Technology Data Exchange (ETDEWEB)

    Gobert, G

    1962-07-01

    Developments for gas-lubricated bearings are discussed. The study of how the fluid film behaves compared to what one can expect is discussed. Various devices are described which have allowed us to go up to peripheral velocity greater than 400 m/s, leading to rotational speed of several thousand revolutions per second. This is obtained by using an automatic setting for the distance between fixed and rotating walls. (author) [French] Le present travail debute par un apercu de l'etat actuel des connaissances sur le processus de la lubrification, notamment sur les paliers a lubrification par gaz. Apres une analyse effectuee au moyen des recentes theories de R. Comolet sur le film fluide entre disques paralleles, nous decrivons des machines dont les charges axiales et radiales specialement etudiees ont permis, au moyen d'un reglage automatique des jeux entre parois fixes et parois tournantes, d'atteindre des vitesses peripheriques superieures a 400 m/s correspondant a des vitesses de plusieurs milliers de tours par seconde. (auteur)

  5. Insights gained from PSAs of French 900MWe and 1300MWe units

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.-M.; Villemeur, A.; Berger, J.-P.; Guio, J.-M. de

    1991-01-01

    The two probabilistic safety assessments of 900MWe and 1300MWe Pressurized Water Reactors (PWRs) recently completed in France constitute an important knowledge resource for the assessment of PWR safety. One innovative feature of this research programme, which yielded many valuable lessons, comes from the fact that plant shutdown state and long term post-accident conditions were fully taken into account. (author)

  6. Monitoring-control of the 900 MWe and 1300 MWe nuclear reactors

    International Nuclear Information System (INIS)

    Meyer, J.

    1982-01-01

    After a short definition of the monitoring-control of the 900 MWe and 1300 MWe nuclear reactors, and a recall of requirements of nuclear energy, this paper presents the following points concerning the whole system of monitoring-control: the organization, the systems (instrumentation, automation), the technologies, the imperfections and the improvements brought to the system [fr

  7. CAREM project 15 to 150 MWe

    International Nuclear Information System (INIS)

    1992-05-01

    The main goal of the Carem Project is the introduction of a Inherent Safe Nuclear Power Reactor in the range of low power (15 to 150 MWe). For this low-power application, light-water and low enriched uranium was selected, since using those concepts permits to take full advantage of the special characteristics of low power reactors. INVAP has been involved in the last years in the design and construction of a Carem Reactor, which could cover a range up to 150 MWe, using a multiple-unit approach. It would furnished the 150 MWe, using six Carem Reactors, of proper power, which would share most of the services. INVAP is a reliable supplier of not only the nuclear reactor but also of the fuel

  8. Etude du point critique des paliers lisses alimentés à la graisse Study of Critical Point on Plain Bearings Lubricated by Grease

    Directory of Open Access Journals (Sweden)

    Delneuville P.

    2006-11-01

    Full Text Available Cet article est essentiellement consacré à l'étude du fonctionnement limite des paliers lisses lubrifiés à la graisse ; son but est de permettre la détermination des conditions nécessaires à l'obtention d'un film lubrifiant d'épaisseur suffisante. L'auteur cherche à cette occasion une loi théorique qui permet de départager les régimes de lubrification, non plus empiriquement, mais sur la base de lois statistiques. La formulation finale doit permettre aux utilisateurs de situer correctement les conditions de fonctionnement d'un palier lisse et d'en estimer la sécurité du régime. This article deals essentially with the boundary behaviour of plain bearings lubricated by grease. Its aim is to determine the conditions required to obtain a sufficiently thick lubricating film. The author proposes a theoretical law for separating lubrication types. This law is not empirical but is based on statistical laws. The final formulation should enable users to correctly situate the operating conditions of a plain bearing and to evaluate the safety factor during running.

  9. French 900 MWe PWR PSA preliminary results

    International Nuclear Information System (INIS)

    Lanore, J.M.; Brisbois, J.

    1988-10-01

    A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

  10. Water treatment for 500 MWe PHWR plants

    International Nuclear Information System (INIS)

    Vasist, Sudheer; Sharma, M.C.; Agarwal, N.K.

    1995-01-01

    Large quantities of treated water is required for power generation. For a typical 500 MWe PHWR inland station with cooling towers, raw water at the rate of 6000 m 3 /hr is required. Impurities in cooling water give rise to the problems of corrosion, scaling, microbiological contamination, fouling, silical deposition etc. These problems lead to increased maintenance cost, reduced heat transfer efficiency, and possible production cut backs or shutdowns. The problems in coastal based power plants are more serious because of the highly corrosive nature of sea water used for cooling. An overview of the cooling water systems and water treatment method is enumerated. (author). 2 refs., 1 fig

  11. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  12. Reactor protection systems of 500 MWe PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

    1997-03-01

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

  13. Towards commercial fast breeder reactors the first 1200 MWe unit

    International Nuclear Information System (INIS)

    Banal, M.; Carle, R.

    The public probably thinks of these fast breeder reactors in terms of their rising unit capacity: RAPSODIE 20 MW (thermal), raised to 40 MW, PHENIX 25 MWe, and now 1200 MWe. However, the purposes of the project and the framework of construction have been fundamentally different in each case. Design parameters and the development program of the LMFBR are presented. (auth)

  14. Safety options for the 1300 MWe program

    International Nuclear Information System (INIS)

    Cayol, A.; Dupuis, M.C.; Fourest, B.; Oury, J.M.

    1980-04-01

    Standardization of the nuclear plants built in France implies an examination of the main technical safety options to be taken for a given type of reactor. By this procedure the subjects for which detailed studies will be needed to confirm the decisions made for the project can be defined in advance. In this context the technical safety option analysis for the 1300 MWe plants was conducted from the end of 1975 to the middle of 1978 according to usual regulation examination practice. The main conclusions are presented on the following subjects: safety methods; technical options concerning the containment vessel, primary fluid activity, fuel elements, steam generators; general organization of the lay-out [fr

  15. Paluel: the first of the 1300 MWe class

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The 1300 MWe class follows that of 900 MWe class. It is the result of studies which have taken into account the evolution of projects made by manufacturers, of research into economies of scale and site optimisation and the attempt to secure a reputation in the export field. In comparison with the 900 MWe class, the 1300 MWe class offers both similarities and differences. Similarities: the general design of the pressure vessel and the fuel elements is the same, as is the design of the loops on the primary cooling circuit. With the aim of reducing costs, the Equipment department carried out a study in 1978 regarding a number of slight modifications in the design called P'4, consisting of at least 14 units, orders for which will be given in the period up to 1983-84 [fr

  16. The THESEUS project -- 50 MWe solar thermal power for Crete

    Energy Technology Data Exchange (ETDEWEB)

    Schillig, F.; Geyer, M.; Kistner, R.; Aringhoff, R.; Nava, P.; Brakmann, G.

    1998-07-01

    A consortium of European industry, utilities and research institutions from Greece, Germany, Spain and Italy attempts to implement a 52 MWe solar thermal power plant with parabolic trough technology on the Greek island of Crete sponsored by the EU' s THERMIE program. The increased demand for electricity on the island, a consequence of the growing allurement of the island as a tourist resort, makes it necessary to expand the installed capacity on Crete during the next years. According to the capacity expansion plans of Greek' s utility PPC a 160 MWe heavy fuel-fired power plant complex--two 30 MWe diesel units and two 50 MWe steam turbine units--is foreseen to be built by the year 2002. In this paper a description of the technical, economical and environmental aspects of the THESEUS project is provided. Moreover a market entry strategy for solar thermal power generation is discussed.

  17. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1976-03-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by NSSS supply. (M.S.)

  18. Power distribution monitoring and control in 500 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, A.

    1996-01-01

    The 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) is expected to be commissioned in a few years. It has a relatively large sized core with complex material distribution in comparison to the currently operating 220 MWe PHWRs. The resulting neutronically loosely coupled system demands continuous control of the core power distribution. This paper gives a brief description and analysis of the reactor monitoring and control system proposed for this reactor. (author). 11 refs, 8 figs, 3 tabs

  19. 300 MWe Burner Core Design with two Enrichment Zoning

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang Ji; Kim, Yeong Il

    2008-01-01

    KAERI has been developing the KALIMER-600 core design with a breakeven fissile conversion ratio. The core is loaded with a ternary metallic fuel (TRU-U-10Zr), and the breakeven characteristics are achieved without any blanket assembly. As an alternative plan, a KALIMER-600 burner core design has been also performed. In the early stage of the development of a fast reactor, the main purpose is an economical use of a uranium resource but nowadays in addition to the maximum utilization of a uranium resource, the burning of a high level radioactive waste is taken as an additional interest for the harmony of the environment. In way of constructing the commercial size reactor which has the power level ranging from 800 MWe to 1600 MWe, the demonstration reactor which has the power level ranging from 200 MWe to 600 MWe was usually constructed for the midterm stage to commercial size reactor. In this paper, a 300 MWe burner core design was performed with purpose of demonstration reactor for KALIMER-600 burner of 600 MWe. As a means to flatten the power distribution, instead of a single fuel enrichment scheme adapted in design of KALIMER-600 burner, the 2 enrichment zoning approach was adapted

  20. Instrumentation of steam cycle HTR's up to 900 MWe

    International Nuclear Information System (INIS)

    Leithner, D.E.; Winkenbach, B.

    1982-06-01

    Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

  1. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  2. Experience on KKNPP VVER 1000 MWe water chemistry

    International Nuclear Information System (INIS)

    Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

    2015-01-01

    Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

  3. The french 900 MWe PWR PSA results and specificities

    International Nuclear Information System (INIS)

    Lanore, J.M.

    1990-01-01

    A probabilistic Safety Assessment has been performed by the Safety Analysis Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the level 1 results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures and the importance of risk during shutdown situations

  4. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1975-06-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed alphabetically. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months) (M.S.)

  5. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1975-12-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by country. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months). (M.S.)

  6. Improved design on Qinshan 300 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Shi Peihua; Cheng Wanli; Lu Rongliang

    1993-01-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

  7. Improved design on Qinshan 300 MWe nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

    1993-06-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

  8. A multinode digital control system for 500 MWe PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Patil, G N; Suresh Babu, R M; Jangra, L R; Das, Shantanu; Mallik, S B [Bhabha Atomic Research Centre, Bombay (India). Reactor Control Div.

    1994-12-31

    A fault tolerant distributed digital computer system for 500 MWe reactor power regulation is configured around standard microcomputer boards designed indigenously. The system is configured as functionally partitioned distributed control system having 8 nodes linked by high-speed dual redundant high-way. The paper gives the details of the configuration of system and how the features of fault-tolerance and fail-safeness are achieved through design. (author). 1 fig.

  9. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

  10. Forbush decreases observed 40 mwe underground in 1978

    International Nuclear Information System (INIS)

    Benko, G.; Kecskemety, K.; Neuprandt, G.; Somogyi, A.J.

    1982-01-01

    Forbush decreases observed 40 mwe underground at Budapest in the first half of 1978 have been analysed together with the data of several neutron monitor stations in Europe. Assuming a power-exponential type spectrum for the variations spectrum in space as a function of rigidity, the best fitting values of power and upper cut-off rigidity have been calculated from maximum decrement by means of the weighted least squares method

  11. Supplementary shutdown system of 220 MWe standard PHWR in India

    International Nuclear Information System (INIS)

    Muktibodh, U.C.

    1997-01-01

    The design objective of the shutdown system is to make the reactor subcritical and hold it in that state for an extended period of time. This objective must be realised under all anticipated operational occurrences and postulated abnormal conditions even during most reactive state of the core. PHWR design criteria for shutdown stipulates requirement of two independent diverse and fast acting shutdown systems, either of which acting alone should meet the above objectives. This requirement would normally call for a large number of reactivity mechanism penetrations into the calandria. From the point of view of space availability at the reactivity mechanism area on top of calandria, for the relatively small core of 220 MWe PHWRs, and ease of maintenance realisation of the total worth by either of the shutdown systems acting alone was difficult. To overcome this engineering constraint and at the same time to satisfy the design criteria, a unique approach to meet the reactivity demands for shutdown was adopted. The reactivity requirements of the shutdown consists of fast and slow reactivity changes. For the shutdown system of 220 MWe PHWRs, the approach of realizing fast reactivity changes with dual redundant, diverse, fast acting shutdown systems aided by a slow acting shutdown system to counter delayed reactivity changes was conceived. The supplementary slow acting shutdown system is called upon to act after actuation of either of the two redundant fast acting systems and is referred to as Liquid Poison Injection System (LPIS). The system adds bulk amount of neutron poison (boric acid), equivalent to 45 mk, directly into the moderator through two nozzles in calandria using pneumatic pressure. This paper describes the design of LPIS as envisaged for the standardised 220 MWe PHWRs. (author)

  12. Dynamic response of domes in CANDU 600 MWe containments

    International Nuclear Information System (INIS)

    Aziz, T.S.; Meng, V.; Alizadeh, A.

    1981-01-01

    CANDU reactors of the 600 MWe type are typically housed in a cylindrical prestressed concrete containment structure; rising from a flat slab and ending in a domed roof. The principal components of this structure are: (a) a circular base slab, (b) a vertical cylinder and (c) a spherical dome cap. A unique feature of a CANDU 600 MWe containment structure is the existence of an inner spherical concrete dome, located below the outer spherical dome, which serves as the bottom of a reservoir for the storage of 560,000 imperial gallons of douzing water. The thickness of the prestressed cylinder wall is approximately doubled between the two domes to create a ring beam. Inside the containment there exists an internal concrete structure which is independent of the containment structure except for support on the base slab. The containment boundary is a fully prestressed concrete structure. This paper deals with the seismic behaviour of the CANDU 600 MWe containment structure and the effect of its unique features; such as the lower dome and the douzing water on this behaviour. The objective of the study is to evaluate the interaction (coupling) effects between the different components of the structure. The approach taken is to study each component of the structure individually, then an assembly of the different components, and finally the total containment structure. This presentation is limited to the vertical response of the structure under a vertical earthquake only. Axisymmetric finite elements were used in all models. The vertical responses at selected points of the structure were obtained by the response spectrum method as well as the time-history method. It was observed that the response spectrum method over-estimates the vertical response of the domes and under-estimates the vertical responses of the ring girder and the containment cylinder compared to the time-history method. (orig./RW)

  13. A 600 MWe advanced PWR for the 1990's

    International Nuclear Information System (INIS)

    Lemon, J.E.; Malloy, J.D.; Allen, R.E.

    1987-01-01

    The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) have prepared a conceptual design of an advanced 600 MWe Presurized Water Nuclear Power Plant. This design utilizes the large body of design and operating experience on PWRs in the U.S. and abroad and incorporates improvements emphasizing simplicity, safety, licensability, ease of construction, operability, reliability and maintainability. Cost and schedule estimates based on U.S. utility experience indicate that this plant design should be competitive with alternate options

  14. The status of safeguarding 600 MW(e) CANDU reactors

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

    1982-09-01

    There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

  15. Technical feasibility study of 60 MWe fast reactor concept: RAPID

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

    1993-01-01

    A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

  16. Flux mapping algorithm (FMA) for 700 MWe PHWR

    International Nuclear Information System (INIS)

    Sonavani, Manoj; Ingle, V.J.; Singhvi, P.K.; Raj, Manish; Fernando, M.P.S.; Kumar, A.N.

    2012-01-01

    For large reactor like 700 MWe PHWR effective spatial control is essential and is provided by RRS. For spatial control purpose reactor core is divided into 14 power zones. Corresponding to each zone is a light water zonal compartment. The 14 ZCCs are located in two radial planes, each containing 7 ZCCs. For each zone, power measurement is carried out using inconel (3 pitch long) self powered neutron detector (SPND) at appropriate location close to the respective ZCC. Since the zone power as obtained by the healthy zone control detector (ZCD) reading belonging to a particular zone may not correspond to its actual power because the detector per zone, measure only average fluxes but the zone extends over a large core region. Therefore accurate estimation of zone power calibration factors is required to estimate the zone powers and also to provide effective spatial power control to avoid the xenon induced spatial power oscillations in large PHWRs like 700 and 540 MWe Reactors. This accurate calculation of zone power is carried out by FMS which uses λ modes in its algorithm. Flux at any point inside the reactor can be represented in terms of the linear combination of these modes. Coefficients used in the expansion are called combining coefficient. If the readings of the detectors are known, then combining coefficients can be estimated by simple matrix operations. Once these combining coefficients are known, flux at any point inside the reactor can be found. (author)

  17. Improving 900 MW(e) PWR control rooms

    International Nuclear Information System (INIS)

    Bouat, M.; Marcille, R.

    1983-01-01

    Analyses of the behaviour of operators during operating tests on PWR units and the lessons learned from the TMI-2 accident have demonstrated the need to improve the interface between operators and the facilities they control. To that end, and to complement its establishment of safety panels, Electricite de France (EDF) embarked upon a study on the ''Modification of Control Desks and Boards'' in control rooms. This study, involving twenty-eight 900 MW(e) units, almost all of which are currently in service, began with an ergonomic analysis of control rooms by an external consultant, the ADERSA GERBIOS Association. This analysis was based on interviews with simulator instructors and operators, a study of the operation of the unit, and a general review of previous studies. The analysis began in October 1980 and resulted, in April 1981, in a critical report and a proposal to create a full-scale mock-up of a 900 MW(e) control room. Improvements to this were subsequently proposed, enabling options to be made between, among other things, active overall control panels and function-by-function control panels. Finally, a number of general principles, which largely encompass the operators' suggestions, were defined. The alterations to be made will make it necessary to revamp the control panels completely. The work and tests involved should match the duration of refuelling shut-downs. Audio-visual training programmes are planned (portable model). (author)

  18. Qinshan 300Mwe NPP full scope simulator upgrade

    International Nuclear Information System (INIS)

    Qi Kelin; Li Qing; Liu Wei, Lai Shengyuan

    2006-01-01

    On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

  19. Evolution of MMI for 500 MWe PHWR plant

    International Nuclear Information System (INIS)

    Surendar, Ch.; Sharma, M.P.; Jayanthi, S.

    1994-01-01

    The Indian nuclear power programme for building Pressurized Heavy Water Reactors began with the construction of two units at Kota, Rajasthan. Although the concept of a centralized control room has been used since the beginning, the man-machine interface design has evolved with technological developments. The man-machine interaction in the earliest plants imposed a considerable burden on the operators and led to a need for more sophisticated instrumentation. Several microprocessor and computer based systems were identified and developed and many were retrofitted into existing plants providing immediate advantages. This paper traces the evolution of many of these systems and also describes the basis and the architecture for the man-machine interaction scheme in the 500 MWe nuclear power plants currently being designed. (author). 7 refs., 2 figs., 1 tab

  20. Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

    International Nuclear Information System (INIS)

    Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

    1998-07-01

    At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

  1. Use of gadolinium as neutron poison in 540 MWe PHWR

    International Nuclear Information System (INIS)

    Nag, P.K.; Fernando, M.P.S.; Kumar, A.N.

    2006-01-01

    In Pressurised heavy water reactors (PHWRs), neutron poison in the moderator is used to compensate the excess reactivity present in the core on different occasions such as xenon decay during synchronization just after poison out period or start ups from xenon free conditions. It is also used in secondary shutdown system (SDS-2), where required amount of neutron poison is injected directly into the moderator within 2.5 seconds. Further, it is also used for over poisoning the moderator to achieve the guaranteed shutdown state when the regular shutdown systems are taken for maintenance. Generally, two types of moderator poisons are used in power reactors to balance the reactivity of the core and they are boron and gadolinium. Gadolinium is used in the form of gadolinium nitrate (Gd(NO 3 ) 3 .6H 2 O). The paper gives the details of estimation of reactivity coefficients of gadolinium for 540 MWe PHWR for different operating conditions. These neutron poisons are converted into non-absorbing elements and therefore their effective worth will decrease as reactor operation proceeds. The rate of burning of neutron absorbing isotopes depends on its magnitude of absorption cross-section and thermal flux seen by them. The present study discusses the burning characteristics of gadolinium during power operation in 540 MWe PHWR. It is established by detailed analysis that the rate of positive reactivity realized due to burning of neutron absorbing Gd isotopes almost match with the build up rate of xenon. The burning half lives of boron and gadolinium is worked out for different power levels. (author)

  2. Three-dimensional studies of the 700 MWe steam generator design

    International Nuclear Information System (INIS)

    John, B.; Pietralik, J.

    2006-01-01

    The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of 540 MWe IPHWRs, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG is of the same type and has the same tube bundle design and the same heat transfer area. The tube diameter, tube pitch, and outer diameter of the SG sections are the same as for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer and the flow distribution plate are different in the two designs. The changes were required due to a 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated against the 540 MWe SG data. The analysis was made for the 700 MWe SG for two versions: with and without integral preheater. The results of the THIRST runs were used for a flow-induced vibration analysis. The results of the flow-induced vibration analysis show that the vibrations are not excessive. (author)

  3. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

    Kang, K. M.; Suh, K. Y.

    2010-01-01

    The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

  4. Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

    International Nuclear Information System (INIS)

    Mattei, J.M.; Bars, G.

    1987-11-01

    During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

  5. Study to use graded cobalt adjuster in 540 MWe PHWR

    International Nuclear Information System (INIS)

    Raj, Manish; Fernando, M.P.S.; Pradhan, A.S.; Kumar, A.N.

    2007-01-01

    Full text: There are 17 adjusters in 540 MWe PHWR, which are essentially provided for xenon override function. They also provide flux flattening being in the central region of the reactor core. The present design of adjusters consists of stainless steel tube. The adjuster rods are grouped into 8 banks for movement. Since adjusters are normally fully inserted during reactor operation, they are best suited for production of cobalt 60. The nickel-plated cobalt in the form of either slugs or pellet are used for the design of cobalt pencils. The number of pencils can be varied to optimize the reactivity load and cobalt 60 production requirement. The worth and activity of cobalt adjusters have been worked out considering different pin configuration for the adjuster assembly. To start with we have assumed all adjusters throughout its length are of the same configuration. The flux depression factors within the cobalt pencils have been considered in the estimations of the specific and total cobalt 60 activities. The option of using graded cobalt adjusters, where different pin configuration along the length is considered for better flux flattening

  6. Thermoeconomic Modeling and Parametric Study of Hybrid Solid Oxide Fuel Cell â Gas Turbine â Steam Turbine Power Plants Ranging from 1.5 MWe to 10 MWe

    OpenAIRE

    Arsalis, Alexandros

    2007-01-01

    Detailed thermodynamic, kinetic, geometric, and cost models are developed, implemented, and validated for the synthesis/design and operational analysis of hybrid solid oxide fuel cell (SOFC) â gas turbine (GT) â steam turbine (ST) systems ranging in size from 1.5 MWe to 10 MWe. The fuel cell model used in this thesis is based on a tubular Siemens-Westinghouse-type SOFC, which is integrated with a gas turbine and a heat recovery steam generator (HRSG) integrated in turn with a steam turbi...

  7. Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Colas, A. F. [Commissariat a l' Energie Atomique, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, CEA/IPSN, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, B.P. No. 6, 92260 Fontenay-aux-Roses (France); Morzelle, C. [Service Etudes et Projets Thermiques et Nucleaires, EdF Lyon (France)

    1986-02-15

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

  8. Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

    International Nuclear Information System (INIS)

    Colas, A.F.; Morzelle, C.

    1986-01-01

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

  9. Some failures of diesel generators during commissioning tests of 1300 MWe PWR

    International Nuclear Information System (INIS)

    Colas, A.F.; Morzelle, C.

    1985-10-01

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: Alarm sensors on fuel, lubricating, cooling circuits; Injection pumps and speed governors; Fuel delivery; Vibrations of fuel and lubrication lines. This paper shows how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects

  10. IRIS-50. A 50 MWe advanced PWR design for smaller, regional grids and specialized applications

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Carelli, Mario; Conway, Larry; Hundal, Rolv; Barbaso, Enrico; Gamba, Federica; Centofante, Mario

    2009-01-01

    IRIS is an advanced, medium-power (1000 MWt or ∼335 MWe) advanced PWR design of integral configuration, that has gained wide recognition due to its innovative 'safety-by-design' safety approach. In spite of its smaller size compared to large monolithic nuclear power plants, it is economically competitive due to its simplicity and advantages of modular deployment. However, the optimum power level for a class of specific applications (e.g., power generation in small regional isolated grids; water desalination and biodiesel production at remote locations; autonomous power source for special applications, etc.) may be even lower, of the order of tens rather than hundreds of MWe. The simple and robust IRIS 335 MWe design provides a solid basis for establishing a 20-100 MWe design, utilizing the same safety and economics principles, so that it will retain economic attractiveness compared to other alternatives of the same power level. A conceptual 50 MWe design, IRIS-50, was initially developed and then assessed in a 2001 report to the US Congress on small and medium reactors, as a design mature enough to have deployment potential within a decade. In the meantime, while the main efforts have focused on the 335 MWe design completion and licensing, parallel efforts have progressed toward the preliminary design of IRIS-50. This paper summarizes the main IRIS-50 features and presents an update on its design status. (author)

  11. Obligations and characteristics applicable to the French unit of the 1400 MWe series. Adaptation to the 900 and 1300 MWe series

    International Nuclear Information System (INIS)

    Conte, M.

    1985-10-01

    This report presents the directives concerning the obligations and the main characteristics of the nuclear PWR units of 1400 MWe, notified Electricite de France on the 06th of October 1983 by the Industry and Research Department. They reflect the concept of defence in depth [fr

  12. Determination of plateau slope and activity using filter measurement results and W. Chauvenet's criterion (Mage II and Fortran IV calculation programmes); Determination de pente de palier et d'activite a partir de resultats de mesure filtres selon le critere de W. Chauvenet (programmes de calcul en Mage II et Fortran IV)

    Energy Technology Data Exchange (ETDEWEB)

    Becker, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires, Departement d' Electronique Generale, Laboratoire de Metrologie de la Radioactivite

    1967-10-01

    In order to permit (at least partially) the objective comparison of plateau characteristics of gas-ionisation counters, plateaus which are obtained when each radioactive sample from normal LMR production is measured, a programme has been drawn up (on an electronic computer) comprising a processing section using least squares for obtaining the corrected plateaux and energies. With a view to an automatic operation of radio-measurement chains, the programme also comprises a preliminary section in which the statistical considerations of B. Peirce have been applied in the version better known as Chauvenet's criterion; this has been done with a view to eliminate measurement results which are dubious and even totally wrong. (author) [French] Pour rendre possibles (au moins partiellement) des comparaisons objectives entre paliers de caracteristiques de compteurs a ionisation gazeuse, paliers traces lors de la mesure de chaque etalon radioactif de la production courante du L.M.R., il a ete ecrit un programme (sur machine a calculer electronique) comportant une partie de traitement par les moindres carres en vue de la determination de pentes et d'activites corrigees. En prevision d'une exploitation automatique de cha es de radio-mesure, le programme comporte en outre une partie preliminaire dans laquelle des considerations statistiques dues a B. Peirce ont ete appliquees dans leur version plus connue sous le nom de critere de Chauvenet et ce dans le but d'une elimination des resultats de mesure suspects et meme veritablement aberrants. (auteur)

  13. Domestic design and validation of natural circulation steam generator of China 1000 MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu, H.Y.; Wang, X.Y.; Wu, G.; Qin, J.M.; Xiong, Ch.H.; Wang, W.; Chen, J.L.; Cheng, H.P.; Zuo, Ch.P.

    2005-01-01

    In order to meet the requirements of domestic design of China intending built NPP projects, Research Institute of Nuclear Power Operation (RINPO) has achieved design of 1000 MWe NPP steam generator, called RINSG-1000(means 1000MWe SG designed by RINPO), which is based on SG research ,experiments and service experience accumulated by RINPO in more 40 years. Testing validation of two steam generator key technologies, advanced moisture separate device and sludge collector, has been accomplished during the period of 2000 to 2002. This paper describes the design features of RINSG-1000, and provides some validation test results. (authors)

  14. 1300°F 800 MWe USC CFB Boiler Design Study

    Science.gov (United States)

    Robertson, Archie; Goidich, Steve; Fan, Zhen

    Concern about air emissions and the effect on global warming is one of the key factors for developing and implementing new advanced energy production solutions today. One state-of-the-art solution is circulating fluidized bed (CFB) combustion technology combined with a high efficiency once-through steam cycle. Due to this extremely high efficiency, the proven CFB technology offers a good solution for CO2 reduction. Its excellent fuel flexibility further reduces CO2 emissions by co-firing coal with biomass. Development work is under way to offer CFB technology up to 800MWe capacities with ultra-supercritical (USC) steam parameters. In 2009 a 460MWe once-through supercritical (OTSC) CFB boiler designed and constructed by Foster Wheeler will start up. However, scaling up the technology further to 600-800MWe with net efficiency of 45-50% is needed to meet the future requirements of utility operators. To support the move to these larger sizes, an 800MWe CFB boiler conceptual design study was conducted and is reported on herein. The use of USC conditions (˜11 00°F steam) was studied and then the changes, that would enable the unit to generate 1300°F steam, were identified. The study has shown that by using INTREX™ heat exchangers in a unique internal-external solids circulation arrangement, Foster Wheeler's CFB boiler configuration can easily accommodate 1300°F steam and will not require a major increase in heat transfer surface areas.

  15. Concept of voltage monitoring for a nuclear power plant emergency power supply system (PWR 1300 MWe)

    International Nuclear Information System (INIS)

    Andrade, R.B. de

    1988-01-01

    Voltage monitoring concept for a Nuclear Power Plant Emergency Power Supply Systems (PWR 1300 MWe) is described based on the phylosophy adopted for Angra 2 and 3 NPP's. Some suggested setpoints are only guidance values and can be modified during plant commissioning for a better performance of the whole protection system. (author) [pt

  16. Main problems experienced on diesel generators of French 900 MWe operating units

    Energy Technology Data Exchange (ETDEWEB)

    Dredemis, Geoffroy; Jude, Francois [Commissariat a l' Energie Atomique, centre d' Etudes Nucleaires de Fontenay-aux-Roses, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

    1986-02-15

    Each unit of all the French nuclear power plant is equipped with two diesel emergency generator sets., For the totality of standards PWRs of 900 MWe, they are identical. We present in this communication the most significative failures met with diesel engines on operating units, such as rupture of fuel injection pipes, breaking of the connecting rods, and cylinder lubrication failures. All these incidents, which affected the emergency power sources of concerned units, had generic characteristics. In view of their potential consequences, it was proceeded in each case to an immediate control of the components concerned of all PWR 900 MWe diesel engines. At the same time, studies were started as to what modifications would permit to solve rapidly each one of the problems met with. (authors)

  17. Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

    International Nuclear Information System (INIS)

    Fath, H.E.S.

    1988-01-01

    Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

  18. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

    Science.gov (United States)

    1981-01-01

    The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

  19. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Design Requirements Document (DRD)

    Science.gov (United States)

    Rigo, H. S.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Bents, D. J.; Hatch, A. M.

    1981-01-01

    A description and the design requirements for the 200 MWe (nominal) net output MHD Engineering Test Facility (ETF) Conceptual Design, are presented. Performance requirements for the plant are identified and process conditions are indicated at interface stations between the major systems comprising the plant. Also included are the description, functions, interfaces and requirements for each of these major systems. The lastest information (1980-1981) from the MHD technology program are integrated with elements of a conventional steam electric power generating plant.

  20. Secondary cycle water chemistry for 500 MWe pressurised heavy water reactor (PHWR) plant: a case study

    International Nuclear Information System (INIS)

    Bhandakkar, A.; Subbarao, A.; Agarwal, N.K.

    1995-01-01

    In turbine and secondary cycle system of 500 MWe PHWR, chemistry of steam and water is controlled in secondary cycle for prevention of corrosion in steam generators (SGs), feedwater system and steam system, scale and deposit formation on heat transfer surfaces and carry-over of solids by steam and deposition on steam turbine blades. Water chemistry of secondary side of SGs and turbine cycle is discussed. (author). 8 refs., 2 tabs., 1 fig

  1. Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, S.

    1982-01-01

    Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

  2. A review of start-up operations on the first units of the 1300 MWe generation

    International Nuclear Information System (INIS)

    Meclot, B.; Lemagny Boc Lonlaygue, C.; Lavogiez, M.

    1986-01-01

    This paper describes and offers comments on the different phases of start-up on power stations of the P4 series. Then, one reviews incidents which occurred in the course of these start-up phases and, having highlighted the lessons to be learnt from the commissioning of these power stations, goes on to make a comparative study of 1300 and 900 MWe availability in the initial year of operation [fr

  3. Evolution of the on-site electric power sources on French 900 MWe PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Bera, Jean [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, Departement d' Analyse de Surete, Service d' Analyse Fonctionnelle, Institut de Protection et Surete Nucleaire, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

    1986-02-15

    Additional means have been provided on the French 900 MWe PWRs to improve safety if both the off-site and on-site Power sources are lost, namely: - a primary pump seal water injection device, one for two units; - a gas turbine generator for each site; - supplying any failing unit with electric power from a house load operating unit; - supplying a unit from a diesel generator of another unit. (author)

  4. Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

    International Nuclear Information System (INIS)

    Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

    2015-01-01

    Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

  5. Reliability investigation for the ECC subsystem of a 1300 MWe-PWR

    International Nuclear Information System (INIS)

    Lalovic, M.

    1983-01-01

    In this study, a fault-tree analysis is used for reliability investigation of Emergency Core Cooling Sub-system of a 1300 MWe pressurised water reactor. Basic assumptions of the study are large break in the reactor coolant system and independence of the pseudo-components. Relatively high non-availability of the sub-system was calculated. Critical component and minimum cut set are determined. (author)

  6. Fire probability safety analysis in France for 900 MWe nuclear power plants

    International Nuclear Information System (INIS)

    Bertrand, R.; Bonneval, F.; Mattei, J.M.

    2000-01-01

    This paper describes the methodology implemented by the Institute for Nuclear Safety and Protection (IPSN) to carry out the Fire Probabilistic Safety Assessment (Fire PSA) for French 900 MWe pressurised water reactors. The initial results obtained are presented. Additional research and development activities are indicated which IPSN carried out or decided to perform in order to reduce the amount of uncertainty associated with the data or to confirm hypotheses that can impact significantly the study results. (orig.) [de

  7. Safety margin improvement by adopting the feature of interleaving in 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, Nrependra; Yadav, S.K.; Khan, T.A.; Dixit, A.; Singhal, Mukesh; Nair, Suma R.

    2015-01-01

    Indian Pressurised Heavy Water Reactors (IPHWRs) of 700 MWe are under construction at Kakrapar Atomic Power Project -3,4 and Rajasthan Atomic Power Project-7,8. These units have enhanced safety features with respect to standard IPHWRs. One of the enhanced features is interleaving of feeders/channels. In interleaved feeder configuration, each header located at either end of reactor gets connected to one quarter of core channels, which are uniformly distributed. The core is divided into two loops with feeder connected in interleaved fashioned. In this paper a comparative study has been performed between the two cases: 1) The core splits in two vertical halves and each vertical half is a loop of PHT (TAPS-3 and 4 Type configuration). 2) The core is divided into two loops with feeders/ channels connected in interleaved fashioned (700 MWe Configuration). LOCA studies have been performed for 700 MWe PHWR considering interleaving of feeders configuration using in-house developed computer code ATMIKA and 3-D neutron kinetics code IQS-3D. The issue of interleaving is closely linked to an inherent reactivity characteristic of PHWR reactors (viz., positive void reactivity coefficient) which leads to a power increase following a Large LOCA. In 700 MWe PHWR with intent to improve the safety margin, adopted the feature of interleaving of feeders which causes in reduction in the magnitude of void coefficient and results in reduction of peak power during LBLOCA. The systematic LBLOCA study demonstrates that interleaved configuration of feeder/channels of two loops has higher safety margins (i.e. with respect to peak power, prompt-criticality margin, adiabatic heat deposition on the fuel pins, sheath temperature excursion and clad oxidation) with regard to the effectiveness of shutdown system. (author)

  8. Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, Sadamu.

    1982-03-01

    Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

  9. Design, development and deployment of special sealing plug for 540 MWe PHWRs

    International Nuclear Information System (INIS)

    Sharma, G.; Roy, S.; Patel, R.J.

    2012-01-01

    The coolant channel in Pressurized Heavy Water Reactors is a pressure boundary component and is very important for reactor performance and reactor safety. Monitoring the condition of the pressure tube of each coolant channel on a periodic basis is very important. In-Service Inspection (ISI) of the coolant channels in water filled condition is done regularly for 220 MWe PHWR. For the same purpose BARC Channel Inspection System is developed for 540 MWe PHWR also. Special Sealing Plug has been developed to facilitate the channel inspection (in water filled condition) with all necessary safety features at par with normal sealing plug. Special Sealing Plug provides a 50 mm through hole for passage of drive tube of Inspection Head maintaining integrity of PHT. Lot of challenges were faced for developing the Special Sealing Plug and its associated tools. It was a first of its kind design. First ISI of TAPS-4 was conducted successfully using this plug along with associated tools in November 2011. This development has provided immense help to NPCIL in life management of 540 MWe PHWR coolant channels. (author)

  10. Capital cost: high and low sulfur coal plants-1200 MWe. [High sulfur coal

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This Commercial Electric Power Cost Study for 1200 MWe (Nominal) high and low sulfur coal plants consists of three volumes. The high sulfur coal plant is described in Volumes I and II, while Volume III describes the low sulfur coal plant. The design basis and cost estimate for the 1232 MWe high sulfur coal plant is presented in Volume I, and the drawings, equipment list and site description are contained in Volume II. The reference design includes a lime flue gas desulfurization system. A regenerative sulfur dioxide removal system using magnesium oxide is also presented as an alternate in Section 7 Volume II. The design basis, drawings and summary cost estimate for a 1243 MWe low sulfur coal plant are presented in Volume III. This information was developed by redesigning the high sulfur coal plant for burning low sulfur sub-bituminous coal. These coal plants utilize a mechanical draft (wet) cooling tower system for condenser heat removal. Costs of alternate cooling systems are provided in Report No. 7 in this series of studies of costs of commercial electrical power plants.

  11. Development of manufacturing process for production of 500 MWe calandria sheets

    International Nuclear Information System (INIS)

    Hariharan, R.; Ramesh, P.; Lakshminarayana, B.; Bhaskara Rao, C.V.; Pande, P.; Agarwala, G.C.

    1992-01-01

    Calandria tubes made of zircaloy-2 are being used as structural components in pressurised heavy water power reactors. The sheets required for producing calandria tube for 235 MWe reactors are being manufactured at Zircaloy Fabrication Plant (ZFP), NFC utilizing a 2 Hi/4 Hi rolling mill procured for the purpose, by carrying out cold rolling process to achieve the required size after hot rolling suitable extruded slabs. Due to limitation of width of the sheet that can be rolled with the mill as well as the size of the slab that can be extruded with the existing press, difficulties arose in producing acceptable full length sheets of size 6600 mm long x 435 mm wide x 1.6 mm thick for manufacturing 500 MWe calandria tube. This paper deals with the details of the process problem resolved. They are: (a)designing of suitable hot and cold rolling pass schedules, (b)selection and standardization of process parameters such as beta quenching, hot rolling and cold rolling, and (c)details of the overall manufacturing process. Due to implementation of above, sheets required for manufacturing 500 MWe calandria tube sheets were successfully rolled. About 40 nos. of acceptable full length sheets have already been manufactured. (author). 1 fig., 3 tabs

  12. Studies on flow induced vibration of reactivity devices of 700 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, K.M., E-mail: kmprabha@yahoo.com [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Goyal, P.; Dutta, Anu; Bhasin, V.; Vaze, K.K.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Pillai, Ajith V.; Mathew, Jimmy [Nuclear Power Corporation of India Ltd., Mumbai 400 094 (India)

    2012-03-15

    Highlights: Black-Right-Pointing-Pointer FIV studies on internals of heavy water filled calandria of 700 MWe Indian PHWR is presented. Black-Right-Pointing-Pointer This includes CFD and structural dynamic analysis to predict the dynamic behavior of component lying inside calandria. Black-Right-Pointing-Pointer Results of these calculations as well as conclusions from this investigation are presented. Black-Right-Pointing-Pointer It is established that FIV is not a concern in the present design of calandria internals. - Abstract: Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

  13. Evaluation of the reliability of the protection system of 1300 MWE PWR'S

    International Nuclear Information System (INIS)

    Blin, A.

    1990-01-01

    An assesment of the reliability of the Digital Integrated Protection System (SPIN) of the 1300 MWe type french reactors has been carried out by treating an example: the emergency shutdown, which can be called upon by several initiating events. The whole chain, from sensors to breakers and control rods, is taken into account. The reliability parameters used for the quantification are evaluated essentially from the experience feedback of french reactors. The not wellknown parameters being the common cause failure rates of electronic components and the efficiency rate of the self-tests, the results of the study are then presented in a parametric form, according to these two factors

  14. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1997-01-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enables faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk represented by deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible. (author)

  15. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  16. Capture of SO2 by limestone in a 71 MWe pressurized fluidized bed boiler

    Directory of Open Access Journals (Sweden)

    Shimizu Tadaaki

    2003-01-01

    Full Text Available A 71 MWe pressurized fluidized bed coal combustor was operated. A wide variety of coals were burnt under fly ash recycle conditions. Limestone was fed to the combustor as bed material as well as sorbent. The emission of SO^ and limestone attrition rate were measured. A simple mathematical model of SO? capture by limestone with intermittent solid attrition was applied to the analysis of the present experimental results. Except for high sulfur fuel, the results of the present model agreed with the experimental results.

  17. Design of a 2.5MW(e) biomass gasification power generation module

    Energy Technology Data Exchange (ETDEWEB)

    McLellan, R.

    2000-07-01

    The purpose of this contract was to produce a detailed process and mechanical design of a gasification and gas clean up system for a 2.5MW(e) power generation module based on the generation of electrical power from a wood chip feed stock. The design is to enable the detailed economic evaluation of the process and to verify the technical performance data provided by the pilot plant programme. Detailed process and equipment design also assists in the speed at which the technology can be implemented into a demonstration project. (author)

  18. Improved identification to prevent transposition during operation of 900 MWe PWR reactors

    International Nuclear Information System (INIS)

    Leckner, J.M.; Dien, Y.; Cernes, A.

    1986-04-01

    Detailed human factors analysis of 900 MWe PWR control room identification systems was carried out by the Nuclear and Fossil Generation Division of Electricite de France (EDF) consequent to a series of incidents where personnel confused one plant unit, room or piece of equipment for another. Preliminary analysis uncovered coding inadequacies and suggested possible remedies. This data was used to prepare specifications for identification redesign at a pilot plant on which detailed investigations could be carried out. Recommended solutions were submitted to pilot plant operators and their opinion sollicited. Operator recommendations will be tried out on the pilot plant and adopted on a grid-wide basis if trials prove satisfactory

  19. On-site control of 900 and 1300 MWe nuclear reactors control rod assemblies

    International Nuclear Information System (INIS)

    Lacroix, R.; Lebuffe, C.; Bour, D.; Pasquier, T.

    1990-01-01

    To measure the external wear of clads of the RCCA rodlets in both 900 and 1300 MWe P.W.R., two on site examination tools was developed by FRAGEMA. They have been used in 42 inspections between 1986 and 1989. The examination is performed in two successive phases: - longitudinal detection of wear by eddy currents, - characterization of wear by ultrasonic profilometry. Moreover, at the instance of E.D.F., an equipment is developing by INTERCONTROLE. These measurement tools allow a suitable monitoring system adapted to the phenomenon kinetics [fr

  20. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1996-01-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible

  1. 'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

    Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

  2. The lateral distribution of muons in showers at 40 mwe underground

    International Nuclear Information System (INIS)

    Bergamasco, L.; Castagnoli, C.; Dardo, M.; D'Ettorre Piazzoli, B.; Mannocchi, G.; Picchi, P.; Visentin, R.; Sitte, K.; Freiburg Univ.

    1975-01-01

    The multiplicity distribution of muon showers at 40 mwe underground was studied with a 4 m 2 spark chamber telescope. The observed frequencies deviate systematically from those calculated with the 'standard' lateral distributions of Vernov or of Greisen. Agreement can be attained if an enhancement of the muon component at small shower sizes is assumed, in accordance with the assumptions of a two-component theory of cosmic ray origin. It is improved by introducing an age dependence of the lateral structure function. (orig.) [de

  3. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G.; Zaleski, C.P. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les

  4. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G; Zaleski, C P [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les projets de reacteurs futurs

  5. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-08-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990

  6. PC based manual and safety logic card test setup for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Chandgadkar, G.M.; Kohli, A.K.; Agarwal, R.G.; Chandra, Rajesh

    1992-01-01

    Fuel handling controls for 235 MWe PHWR make use of Manual and Logic cards (MLCs) for providing safety interlocks. These cards consist of various type of logic blocks. By connecting these logic blocks all the safety interlocks required for fuel handling controls have been provided. Previously trouble shooting of these cards was done by means of logic probe. Since the method was manual, it was laborious and time consuming. PC based test setup has overcome this drawback and detects the fault at the component level within few seconds. It also gives printout of status of faulty MLC cards. Here motherboard has been designed having slots for insertion of MLC cards. The input/output connection of these cards are coming to two 50 pin FRC connectors. PC communicates through 144 line digital input/output card with MLC card under test. Software is user friendly and outputs suitable input patterns to the card under test and checks for output pattern. It compares this output pattern with compare pattern and detects the fault and displays the symptoms. This system is currently in use at test facility for fuelling machine for 235 MWe PHWR reactor at Refuelling Technology Division, Hall-7. This test setup has been proposed for use at NAPP and future reactors. (author). 4 figs., 1 annexure

  7. Role of pressuriser in enhancing pressure control system capability in primary system of 500 MWe PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Walia, M P.S.; Misri, Vijay; Bapat, C N; Sharma, V K [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The primary heat transport system of a pressurized heavy water reactor (PHWR) extracts and transports the heat produced in the fuel (located inside coolant channel assemblies) to the steam generators where steam is generated to run the turbo-generator. The heat transport medium (primary coolant) is heavy water which is kept in a pressurized liquid state with the help of a pressure control system. Feed and bleed circuits with associated equipment of PHT main system have traditionally constituted the pressure control system. However, for large size reactors of 500 MWe capacity, a surge tank known as pressurizer was incorporated due to the presence of relatively large inventory in PHT main circuit. The pressurizer acts as a cushion for pressure variations resulting from various transients. This significantly reduces the onerous demand on feed and bleed system, thereby reducing reactor outages on system pressure excursions. The paper describes in detail the pressure control system of 500 MWe PHWR involving pressuriser and feed and bleed system including their functions and instrumentation. The results of mathematical modelling/analysis undertaken to establish the response adequacy of pressure control system, to postulated plant transients vis-a-vis the role of pressurizer are presented. (author). 10 figs.

  8. Design of shutdown system no.2 liquid poison injection system for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Bhatnagar, S.; Balasubrahmanian, A.K.; Pillai, A.V.

    1997-01-01

    Defence in depth and two group system concepts form the basic design philosophy for the shutdown systems. There are two independent, diverse and fast acting shutdown systems provided for the 500 MWe PHWR. The design is based on fail-safe principle, sufficient component redundancy and on-line testing. Liquid poison injection system, as shutdown system 2, is newly developed for the 500 MWe PHWRs. The system operates by rapidly injecting gadolinium nitrate solution into bulk moderator using stored helium pressure thereby inserting negative reactivity. A high pressure helium supply tank which provides the energy for system actuation, is connected, through an array of fast acting valves in series-parallel arrangement, to the individual poison tanks storing gadolinium nitrate solution. The valves, belonging to three different channels of reactor Protection System 2, are the only active components in the system. The valves are fail safe and are periodically tested on-line without actually firing the system. The system comprising of in-core assemblies and the external process system has been engineered. Experimental work is being carried out by BARC for design validation and data generation. This paper describes the conceptual development, design basis, design parameters and detailed engineering of the system. (author)

  9. ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2009-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

  10. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  11. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-01-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990. (author)

  12. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  13. Adding a much needed 300 MWe at South Africa's Arnot coal fired power plant

    Energy Technology Data Exchange (ETDEWEB)

    Rich, G. [Alstom, Rugby (United Kingdom)

    2008-12-15

    As power stations built in the last thirty years approach the end of their design life, and the cost of new capacity continues to increase, along with demands for improved efficiency and lower emissions, an integrated approach to retrofit looks increasingly compelling. The ambitious upgrade project currently underway at the Arnot coal fired plant in South Africa, which will result in an update from 6 x 350 MWe to 6 x 400 MWe and a life extension of 20 years, illustrates the benefits. 2 figs.

  14. Evaluation of feed and bleed cooling mode in case of total loss of feedwater on 900 MWe PWR

    International Nuclear Information System (INIS)

    Champ, M.; Cornille, Y.

    1989-07-01

    The physical studies carried out with the CATHARE code to assess the feed and bleed procedure developed in order to cope with the total loss of feed water on a 900 MWe PWR are presented. These studies allowed the definition of the maximum delays of intervention which would prevent the core from uncovering. Different cases of equipment availability are considered. The data generated will be used in the 900 MWe Probabilistic Safety Assessment which is under way at the Institut de Protection et de Surete Nucleaire

  15. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  16. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  17. FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

    International Nuclear Information System (INIS)

    Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

    1989-01-01

    Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

  18. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  19. 400-MWe consolidated nuclear steam system (CNSS). 1255 MWt CNSS design/cost update

    International Nuclear Information System (INIS)

    1984-07-01

    Since 1976 Babcock and Wilcox (B and W) has been extensively involved in the development of a medium-sized (1255 MWt/400 MWe) reactor. Under the sponsorship of the U.S. Department of Energy (DOE) and through a contract with Oak Ridge National Laboratories (ORNL), B and W investigated the feasibility of the concept for utility power generation and cogenerated process heat. The potential benefits of the design, called the Consolidated Nuclear Steam System (CNSS), were also identified. This study provides an update of the CNSS design and cost reflecting current regulatory requirements and operating reactor experience. The study was funded by DOE through ORNL and was performed by B and W and UE and C

  20. Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

    International Nuclear Information System (INIS)

    Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

    1986-01-01

    In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

  1. Stresses imposed by coolant channel end shield interaction in 200 MWe PHWR

    International Nuclear Information System (INIS)

    Mehra, V.K.; Singh, R.K.; Soni, R.S.; Kushwaha, H.S.; Kakodkar, A.

    1983-01-01

    End shield of 200 MWe Pressurised Heavy Water Reactor (PHWR) is a composite tube sheet structure consisting of two circular tube sheets joined together by lattice tubes. Each lattice tube houses a coolant channel assembly which is connected to the end shield through shock absorber device. End shield assembly is suspended in the vault by hanger rods and its horizontal position is controlled by a set of pre-compressed springs. Coolant channel assemblies elongate due to their exposure to fast neutron flux in the reactor. This permanent elongation is monitored periodically. When growth of the channel exceeds a present value, it is prevented from further elongation by the shock absorbing device. Resultant force exerted on the end shield makes it move. This paper describes a numerical method used for evaluating these forces and movement of the end shield. Stresses produced by these forces are calculated by using finite element method. Typical stress values are verified by strain gauge measurements. (orig.)

  2. Modification to 200 MW(e) CANDU for improved dynamic behaviour

    International Nuclear Information System (INIS)

    Chamany, B.F.; Murthy, L.G.K.; Ray, R.N.

    1976-01-01

    Rajasthan Atomic Power Station is inherently suitable for base load operation. Its control philosophy is based on turbine following the reactor. However, due to load fluctuations and inherent limitation of the control system, there had been considerable number of outages of the station. This limitation is further enhanced by improper choice of the operating pressure range of the boilers. Besides, existing fuel design does not permit thermal cycling and hence there is no use in attempting to make the reactor follow the turbine. Design modifications have been suggested for incorporation in the further 200 MW(e) systems. The method adopted is complete decoupling of the reactor from the load. Dynamic behaviour of the station with the suggested modifications and its comparison with the existing situation has been brought out. (author)

  3. Current status of 700 MWe class PHWR NSSS design and engineering technology

    International Nuclear Information System (INIS)

    Park, Tae Keun; Suh, Sung Ki

    1996-06-01

    The capability of NSSS design and engineering technology of KAERI for 700 MWe class PHWR (CANDU 6) as of 1996 March 30 is comprehensively summarized in this report. The design and engineering capability of KAERI which have been gained during the implementation of Wolsung 2, 3 and 4 project are assessed, and showed with tangible scale. The status of Technology Transfer Materials received from Atomic Energy of Canada Limited under the Technology Transfer Agreement (TTA) which is effective simultaneously to Wolsung 3 and 4 contract, is also given in this report. The division of responsibility (DOR) of KAERI for Wolsung 2 and Wolsung 3 and 4 contract is also given, and expansion of DOR from Wolsung 2 contract to Wolsung 3 and 4 is presented. 3 refs. (Author)

  4. Use of artificial neural network in estimating channel power distribution of a 220 MWe PHWR

    International Nuclear Information System (INIS)

    Dubey, B.P.; Chandra, A.K.; Govindarajan, G.; Jagannathan, V.; Kataria, S.K.

    1998-01-01

    Knowledge of the distribution of power in all the 306 channels of a Pressurised Heavy Water Reactor (PHWR) as a result of the movement of one or more of the four regulating rods is important from the operation and maintenance point view of the reactor. Conventional computer codes available for this purpose take several minutes to calculate the channel power distribution on PC-AT/486. An Artificial Neural network (ANN), based on the RPROP algorithm has been developed and employed in predicting channel power distribution of a 220 MWe Indian PHWR as a result of a perturbation caused by the movement of one or more of the four regulating rods of the reactor. The ANN based system produces the result of an analysis much faster than that produced by a conventional computer code usually employed for this application. The ANN based system is rugged, accurate and fast, and therefore, has potential to be used in real-time applications. (author)

  5. Systems analysis of a 100-MWe modular liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Morris, E.E.; Rhow, S.K.; Switick, D.M.

    1985-01-01

    The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

  6. Control of the flanges of the thermal barriers fitting the 900 MWe PWR primary pumps

    International Nuclear Information System (INIS)

    Cleurennec, M.; Thebault, Y.; Abittan, E.; Pages, C.; Lhote, P.A.; Randrianarivo, L.

    1998-01-01

    During maintenance visit on 93 D type primary pumps of French 900 MWe nuclear units, cracking has been evidenced on the thermal barrier, first on the flange, on the face of connection of the cooling, water coils, and then on the weld between the housing and the flange. Laboratory examinations have exhibited that this cracking is due to a fatigue phenomenon which is initiated on locations where high residual stresses are present. One pump, in service in a plant, has received an instrumentation in order to determine stress cycling. Measurements of temperature on the surface of the metal have shown the presence of thermal cycling due to the thermohydraulic conditions inside the thermal barrier. A non destructive testing method using ultrasounds has been developed in order to asses the magnitude cracking. Corrective and preventive actions have been implemented for repairing and improving thermal barrier when cracking is detected. (authors)

  7. Co-firing straw and coal in a 150-MWe utility boiler: in situ measurements

    DEFF Research Database (Denmark)

    Hansen, P. F.B.; Andersen, Karin Hedebo; Wieck-Hansen, K.

    1998-01-01

    A 2-year demonstration program is carried out by the Danish utility I/S Midtkraft at a 150-MWe PF-boiler unit reconstructed for co-firing straw and coal. As a part of the demonstration program, a comprehensive in situ measurement campaign was conducted during the spring of 1996 in collaboration...... with the Technical University of Denmark. Six sample positions have been established between the upper part of the furnace and the economizer. The campaign included in situ sampling of deposits on water/air-cooled probes, sampling of fly ash, flue gas and gas phase alkali metal compounds, and aerosols as well...... deposition propensities and high temperature corrosion during co-combustion of straw and coal in PF-boilers. Danish full scale results from co-firing straw and coal, the test facility and test program, and the potential theoretical support from the Technical University of Denmark are presented in this paper...

  8. Estimate of man-rem expenditures for a mature CANDU 600 MW(e) station

    International Nuclear Information System (INIS)

    Kuperman, I.

    1978-08-01

    In recent years, man-rem expenditures at operating stations have come under close scrutiny in order to reduce operating personnel dosage. This increased awareness has led to concerted efforts to improve station design and to improve operating procedures to achieve lower man-rem expenditures. This paper is intended to highlight design improvements that have been made in the CANDU 600 MW(e) design and to show how these improvements will reduce man-rem expenditures. Other considerations, such as station decontaminations of the primary heat transport system and the fuelling machines and stricter chemistry control are presently available to help reduce man-rem consumption. Also, station management operating policy should emphasize man-rem awareness. (author)

  9. Ergonomic design of mosaic control panel and standardised control tile configurations for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Ughade, A.V.; Das, R.N.; Ramakrishnan, S.

    1994-01-01

    A review of control rooms of operating nuclear power plants identified many design problems having potential for degrading the performance of operators. Many indications and controls on existing control panels are placed outside the recommended visual and reach envelopes for acceptable operator usage. As a result, the application of human factor principles was found to be needed. This paper describes the design approach for working out the dimensions of main control room panels and console using human engineering principles and recommends the ergonomic dimensions of the main control room panels and console. Further it gives the basis and works out the control tile configurations for 500 MWe PHWR project. It also suggests the use of a full scale mock up for design evaluation and verification. (author). 7 refs., 4 figs

  10. Design Evaluation of UIS and In-vessel Fuel Transfer Machine for a 1200MWe SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Kim, Seok Hoon; Park, Chang Gyu; Lee, Su Yeon

    2008-11-15

    The report describes the structural applicability of the upper internal structure (UIS) and the in-vessel fuel transfer machine for a 1200MWe sodium cooled fast reactor (SFR) of a pool type. In the conceptual design, a two rotating plug type as a refueling system is considered. For the two rotating plug type, the diameters of large and small rotating plugs are determined by 7.2m and 5.6m, respectively. Through the use of an inner fuel transfer machine and the lift change machine with a fixed arm length of 1.10m installed on a small rotating plug, all the core assemblies are accessed within 7mm accuracy. The UIS diameter is determined by 4.7m, which includes the all control drive lines in upper part, the diameter of UIS lower part is restricted by 4.4 m to secure the rotation angle of a refueling machine.

  11. Considerations in providing purification flows for 500 MWe PHWR primary circuits

    International Nuclear Information System (INIS)

    Sharma, A.K.; Goswami, S.; Bapat, C.N.; Sharma, V.K.

    1995-01-01

    The purpose of the purification system is to keep the primary heat transport (PHT) system clean by removing traces of impurities arising due to corrosion of the carbon steel pipes and heat transfer surfaces and erosion/corrosion of valve trims, pipes and mechanical seals or due to presence of soluble or insoluble fission products. These impurities are undesirable because they are usually radioactive, either naturally or through activation by the neutron flux as they are carried by the coolant through the reactor core. The purification system minimizes the probability of generation of radioactive impurities by controlling the chemistry of PHT coolant so that corrosion is minimum. Various considerations for providing the requisite purification flow to fulfill the above functions for a typical 500 MWe PHWR are presented. (author). 4 refs., 2 tabs., 2 figs

  12. The Clean Coal Technology Program 100 MWe demonstration of gas suspension absorption for flue gas desulfurization

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, F.E.; Hedenhag, J.G. [AirPol Inc., Teterboro, NJ (United States); Marchant, S.K.; Pukanic, G.W. [Dept. of Energy, Pittsburgh, PA (United States). Pittsburgh Energy Technology Center; Norwood, V.M.; Burnett, T.A. [Tennessee Valley Authority, Chattanooga, TN (United States)

    1997-12-31

    AirPol Inc., with the cooperation of the Tennessee Valley Authority (TVA) under a Cooperative Agreement with the United States Department of Energy, installed and tested a 10 MWe Gas Suspension Absorption (GSA) Demonstration system at TVA`s Shawnee Fossil Plant near Paducah, Kentucky. This low-cost retrofit project demonstrated that the GSA system can remove more than 90% of the sulfur dioxide from high-sulfur coal-fired flue gas, while achieving a relatively high utilization of reagent lime. This paper presents a detailed technical description of the Clean Coal Technology demonstration project. Test results and data analysis from the preliminary testing, factorial tests, air toxics texts, 28-day continuous demonstration run of GSA/electrostatic precipitator (ESP), and 14-day continuous demonstration run of GSA/pulse jet baghouse (PJBH) are also discussed within this paper.

  13. Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

    International Nuclear Information System (INIS)

    Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

    2009-01-01

    The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

  14. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Dhruvanarayana, L.; Gupta, H.; Bharathkumar, M.

    1996-01-01

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D 2 0 hydraulics, H 2 0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of valves

  15. Concept of voltage and frequency monitoring for a nuclear power plant normal power supply system - PWR 1300 MWe

    International Nuclear Information System (INIS)

    Andrade, R.B. de

    1990-01-01

    Voltage and frequency monitoring concept for a Nuclear Power Plant Normal Power Supply System (PWR 1300 MWe) is described based on the phylosophy adopted for Angra 2 and e NPP's. Some suggested setpoints are only guidance values and can be modified during plant commissioning for a better performance of the whole protection system. (author) [pt

  16. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Dhruvanarayana, L; Gupta, H; Bharathkumar, M [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D{sub 2}0 hydraulics, H{sub 2}0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of

  17. Improved NOx emissions and combustion characteristics for a retrofitted down-fired 300-MWe utility boiler.

    Science.gov (United States)

    Li, Zhengqi; Ren, Feng; Chen, Zhichao; Liu, Guangkui; Xu, Zhenxing

    2010-05-15

    A new technique combining high boiler efficiency and low-NO(x) emissions was employed in a 300MWe down-fired boiler as an economical means to reduce NO(x) emissions in down-fired boilers burning low-volatile coals. Experiments were conducted on this boiler after the retrofit with measurements taken of gas temperature distributions along the primary air and coal mixture flows and in the furnace, furnace temperatures along the main axis and gas concentrations such as O(2), CO and NO(x) in the near-wall region. Data were compared with those obtained before the retrofit and verified that by applying the combined technique, gas temperature distributions in the furnace become more reasonable. Peak temperatures were lowered from the upper furnace to the lower furnace and flame stability was improved. Despite burning low-volatile coals, NO(x) emissions can be lowered by as much as 50% without increasing the levels of unburnt carbon in fly ash and reducing boiler thermal efficiency.

  18. Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

    2017-12-15

    This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

  19. Final report on the evolution of supporting conditions for the feeders of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Soni, R.S.; Kushawaha, H.S.; Mahajan, S.C.; Kakodkar, A.; Hariprasad, K.

    1994-01-01

    This report deals with the evolution of generic supporting conditions for the feeders of 500 MWe PHWR based on the analysis and qualification of a few representative feeders. There are 196 different feeder pipe configurations for a total of 748 feeders. The present analysis was aimed at evolving a generalised supporting criteria based on the analysis of some representative feeders. The analysis was carried out for various loadings viz. pressure, temperature, dead weight, operating basis earthquake (OBE), safe shutdown earthquake (SSE) and creep loadings. The analysis for OBE and SSE loadings were carried out using response spectrum method. The effect of spacers between various feeders was modelled using higher damping values than those prescribed in ASME code. Based on the above analyses, generic supporting arrangements for the feeders of various groups have been finalized. This report gives details about the mathematical modelling, the analysis approach, the optimised supporting criteria, finalization of grouping and fixing of boundaries between various groups of feeders. (author). 34 refs., 51 figs., 69 tabs

  20. Mathematical modelling of heat absorption capacity of containment spray system in a 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kota, Sampath Bharadwaj; Ali, Seik Mansoor; Balasubramaniyan, V.

    2015-01-01

    This paper presents a mathematical model for estimating the heat removal by containment spray system in the post Loss of Coolant Accident (LOCA) environment. The procedure involves firstly, the calculation of heat removal rates by droplets of spray dispersed in the air-steam mixture by an appropriate direct contact condensation model accounting for the presence of non-condensable gas (air). Parametric influence of droplet size, ambient pressure and temperature on heat flux is brought out. It was found that the heat flux is inversely proportional to the ambient pressure and diameter. A spray module was subsequently developed and incorporated into an in-house containment thermal hydraulics code. The pressure and temperature transients in a 700 MWe PHWR containment building following a Large Break LOCA was obtained using this code. The efficacy of the spray in condensing the steam is shown by comparing the transients with and without the operation of spray system. Parametric studies are also conducted with respect to droplet size and flow rate of water droplet spray. The details of the investigation are presented and discussed in this paper. (author)

  1. Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Research and Development Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.; Homer, G.B.; Shaber, C.R.; Thurow, T.L.

    1981-11-17

    The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

  2. Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.; Homer, G.B.; Spencer, S.G.; Shaber, C.R.

    1980-05-30

    The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

  3. Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

    International Nuclear Information System (INIS)

    Dwivedi, K.P.; Seth, V.K.

    1994-01-01

    A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

  4. Interpretation of out of line control rod experiments for 1300 MWE PWR

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J.L.; Garcia-Fernandez, L.

    1988-01-01

    The present note summarizes the studies we performed recently in order to search a 2D reconstruction procedure for the 1300 MWE PWR power shape, starting from data coming out from thermocouples placed on several fuel assemblies. In classical PWR design, only a few assemblies are equipped with measurement devices, so that it is necessary to interpolate among measure points in order to obtain a complete coverage of the core. A mathematical approach based on the splitting of the power into a reference steady state nominal shape and some ''influence'' and harmonic functions was chosen. The reference steady state power shape, which corresponds to the full power operating mode, is obtained via direct mobile chamber measurements. The perturbations due to the control rod movements are accounted for by specific ''influence'' functions: moreover, harmonics are used to reconstruct the minor effects due to xenon tilts, rod out of line positions and all actual mechanical and thermohydraulic inhomogeneities. The weighting coefficients of the functions are evaluated by a least square method, starting from the distribution of the deviations among the measurements and the reference values.

  5. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    International Nuclear Information System (INIS)

    Sanatkumar, A.; Jit, I.; Muralidhar, G.

    1996-01-01

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs

  6. Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

    1987-01-01

    One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine rive mechanisms. Some of these holes intersect with each other in the housing end-closers and hence end-closers are reinforced accordingly. This also makes the end-closers nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described. (orig.)

  7. Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

    1987-01-01

    One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine drive mechanisms. Some of these holes intersect with each other in the housing end-closures and hence end-closures are reinforced accordingly. This also makes the end-closures nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described

  8. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    Energy Technology Data Exchange (ETDEWEB)

    Sanatkumar, A; Jit, I; Muralidhar, G [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs.

  9. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  10. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor

    International Nuclear Information System (INIS)

    Vendryes, G.; Zaleski, C.P.

    1964-01-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios (∼1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 Δk/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [fr

  11. IRSN-ANCCLI partnership. IRSN-ANCCLI seminar on decennial inspections of 900 MWe reactors - November 2010

    International Nuclear Information System (INIS)

    Rollinger, Francois; Delalonde, Jean-Claude; Hubert, F.; Paulmaz, X.; Tindillere, M.; Lheureux, Y.; Junker, R.; Sene, M.

    2010-11-01

    This seminar addressed the commitment of local information commissions (CLI) in the analysis and follow-up of the third decennial inspections of the French 900 MWe nuclear reactors. A first session addressed topics directly related to these inspections. Contributions proposed under the form of Power Point presentations by experts and representatives of the IRSN, EDF and CLIs addressed the following issues: safety re-examination of EDF 900 MWe reactors at the occasion of the third decennial inspection, activities of the IRSN related to skill management in nuclear power stations, implementation of the third decennial inspection of the unit 1 of the Fessenheim nuclear power station, the issue of follow-up by a local information commission after a decennial inspection. A second session addressed topics not related to decennial inspections and were proposed by Gravelines and Dampierre local information commissions: analysis of significant safety events, issues of skill management in nuclear power stations

  12. Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Choi, Seok Ki; Kim, Seong O

    2011-01-01

    A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

  13. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

    1982-01-01

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  14. Combustion and NOx emission characteristics of a retrofitted down-fired 660 MWe utility boiler at different loads

    Energy Technology Data Exchange (ETDEWEB)

    Li, Z.Q.; Liu, G.K.; Zhu, Q.Y.; Chen, Z.C.; Ren, F. [Harbin Institute of Technology, Harbin (China)

    2011-07-15

    Industrial experiments were performed for a retrofitted 660 MWe full-scale down-fired boiler. Measurements of ignition of the primary air/fuel mixture flow, the gas temperature distribution of the furnace and the gas components in the furnace were conducted at loads of 660, 550 and 330 MWe. With decreasing load, the gas temperature decreases and the ignition position of the primary coal/air flow becomes farther along the axis of the fuel-rich pipe in the burner region under the arches. The furnace temperature also decreases with decreasing load, as does the difference between the temperatures in the burning region and the lower position of the burnout region. With decreasing load, the exhaust gas temperature decreases from 129.8{sup o}C to 114.3{sup o}C, while NOx emissions decrease from 2448 to 1610 mg/m{sup 3}. All three loads result in low carbon content in fly ash and great boiler thermal efficiency higher than 92%. Compared with the case of 660 MWe before retrofit, the exhaust gas temperature decreased from 136 to 129.8{sup o}C, the carbon content in the fly ash decreased from 9.55% to 2.43% and the boiler efficiency increased from 84.54% to 93.66%.

  15. GTHTR 300 economic calculation with Mini G4ECONS as a basis for generation cost of GTHTR 10 MWe calculation

    International Nuclear Information System (INIS)

    Mochamad Nasrullah; Nurlaila

    2014-01-01

    The government plan to build Experimental Power Reactor (EPR) requires measurable economic assessment. The purpose of the study was to recalculate Gas Turbine High Temperature Reactor of 300 MWe (GTHTR 300) and compare the results with reference data. Then calculate generation cost of GTHTR 3, 5 and 10 MWe using the scale factor calculation. The methodology used is covered the generation cost calculation using the Mini G4Econs spread sheet models published by IAEA. Result of the verification calculation showed that a relatively similar, which means that the calculation model could be used to calculate for same other cases. Afterward, using scale factor, smaller scale reactor could be calculated. The calculation result show that electricity generation cost of SMR-HTR type with load factor 90% and discount rate 10% for power capacity 3, 5 and 10 MWe are 29.5, 22.68 and 16.17 cents$/kWh respectively. However, because the EPR is planning to be built as a non-commercial power reactors, then 5 % and 3 % of discount rate could be chosen, each of those discount rate will result electricity generation cost of 10.37 cents$/kWh and 8.56 cents$/kWh respectively. These result could be considered by the government for developing SMR type of HTR. (author)

  16. STUDY ON DISCHARGE HEAT UTILIZATION OF 250 MWe PCMSR TURBINE SYSTEM FOR DESALINATION USING MODIFIED MED

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-03-01

    Full Text Available PCMSR (Passive Compact Molten Salt Reactor is one type of Advanced Nuclear Reactors. The PCMSR has benefit charasteristics of very efficient fuel use, high safety charecteristic as well as high thermodinamics efficiency. This is due to its breeding capability, inherently safe characteristic and totally passive safety system. The PCMSR design consists of three module, i.e. reactor module, turbine module and fuel management module. Analysis in performed by parametric calculation of the turbine system to calculate the turbine system efficiency and the hat available for desalination. After that the mass and energi balance of desalination process are calculated to calculate the amount of distillate produced and the amount of feed sea water needed. The turbine module is designed to be operated at maximum temperature cycle of 1373 K (1200 0C and minimum temperature cycle of 333 K (60 0K. The parametric calculation shows that the optimum turbine pressure ratio is 4.3 that gives the conversion efficiency of 56 % for 4 stages turbine and 4 stages compressor and equiped with recuperator. In this optimum condition, the 250 MWe PCMSR turbine system produces 196 MWth of waste heat with the temperature of cooling fluid in the range from 327 K (54 0C to 368 K (92 0C. This waste heat can be utilized for desalination. By using MMED desalination system, this waste heat can be used to produce fresh water (distillate from sea water feed. The amount of the destillate produced is 48663 ton per day by using 15 distillation effects. The performance ratio value is 2.8727 kg/MJ by using 15 distillation effects. Keywords: PCMSR, discharged heat, MMED desalination   PCMSR (Passive Compact Molten Salt Reactor merupakan salah satu tipe dari Reaktor Nuklir Maju. PCMSR memiliki keuntungan berupa penggunaan bahan bakar yang sangat efisisien, sifat keselamatan tinggi dan sekaligus efisiensi termodinamika yang tinggi. Hal ini disebabkan oleh kemampuan pembiakan bahan bakar, sifat

  17. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  18. Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2015-02-15

    Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  19. Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

    Science.gov (United States)

    Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

  20. Effect of flow configuration on moderator temperature distribution for 700 MWe Calandria

    International Nuclear Information System (INIS)

    Bharj, Jaspal Singh; Sahaya, R.R.; Dharne, S.P.

    2009-01-01

    The Calandria of a Pressurized Heavy Water Reactor (PHWR) is essentially a horizontal cylindrical vessel housing a matrix of horizontal tubes called Calandria tubes within which is contained the pressure tubes that house the fuel bundles. In addition there are horizontal and vertical flux control and shutdown devices. The Calandria is filled with heavy water moderator at a pressure slightly above the atmosphere. A large amount of heat (about 125 MWth) is generated within the moderator mainly due to neutron slowing down and attenuation of gamma radiations. This heat generation gives rise to a strong buoyancy-driven natural convection flow. In the proposed configuration of 700 MWe PHWR Calandria, moderator inlet diffusers are directed upwards and the outlet nozzles are at the bottom of the Calandria. The basis for the above said inlet/outlet configuration depends upon the various factors like space availability, NPSH requirement for the moderator pumps, and interference of flow with the other components inside the Calandria. This configuration is not conducive for the buoyancy-dominated flows generated due to large volumetric heat generation in the moderator. In order to see the effects of changes in flow configuration by re-orienting the inlet/outlet, a CFD study was undertaken for moderator flows in the conceptual Calandria. In the study, the moderator inlet diffusers direct the cool moderator towards the bottom of the Calandria and hot moderator flows out through the outlets in the upper half of the Calandria. The results of the study with various flow configurations show that modification in moderator flow configuration in Calandria, by way of introduction of moderator in the downward direction through diffusers and provision of the exits from the upper portion of the Calandria, results in significant reduction of the maximum temperature of moderator in Calandria. Further, the temperature distribution in the Calandria in the proposed configurations is much more

  1. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  2. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

    2003-01-01

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  3. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.

  4. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    International Nuclear Information System (INIS)

    Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

    1997-01-01

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

  5. CONCEPTUAL DESIGN AND ECONOMICS OF A NOMINAL 500 MWe SECOND-GENERATION PFB COMBUSTION PLANT

    Energy Technology Data Exchange (ETDEWEB)

    A. Robertson; H. Goldstein; D. Horazak; R. Newby

    2003-09-01

    Research has been conducted under United States Department of Energy Contract DE-AC21-86MC21023 to develop a new type of coal-fired plant for electric power generation. This new type of plant, called a Second Generation Pressurized Fluidized Bed Combustion Plant (2nd Gen PFB), offers the promise of efficiencies greater than 48 percent, with both emissions and a cost of electricity that are significantly lower than those of conventional pulverized coal-fired (PC) plants with wet flue gas desulfurization. The 2nd Gen PFB plant incorporates the partial gasification of coal in a carbonizer, the combustion of carbonizer char in a pressurized circulating fluidized bed boiler, and the combustion of carbonizer syngas in a gas turbine combustor to achieve gas turbine inlet temperatures of 2300 F and higher. A conceptual design and an economic analysis was previously prepared for this plant. When operating with a Siemens Westinghouse W501F gas turbine, a 2400psig/1000 F/1000 F/2-1/2 in. Hg. steam turbine, and projected carbonizer, PCFB, and topping combustor performance data, the plant generated 496 MWe of power with an efficiency of 44.9 percent (coal higher heating value basis) and a cost of electricity 22 percent less than a comparable PC plant. The key components of this new type of plant have been successfully tested at the pilot plant stage and their performance has been found to be better than previously assumed. As a result, the referenced conceptual design has been updated herein to reflect more accurate performance predictions together with the use of the more advanced Siemens Westinghouse W501G gas turbine. The use of this advanced gas turbine, together with a conventional 2400 psig/1050 F/1050 F/2-1/2 in. Hg. steam turbine increases the plant efficiency to 48.2 percent and yields a total plant cost of $1,079/KW (January 2002 dollars). The cost of electricity is 40.7 mills/kWh, a value 12 percent less than a comparable PC plant.

  6. Improvements on computerized procedure system of advanced power reactor 1400 MWe

    International Nuclear Information System (INIS)

    Seong, Nokyu; Jung, Yeonsub; Sung, Chanho; Kang, Sungkon

    2017-01-01

    Plant procedures are instructions to help operator in monitoring, decision making, and controlling Nuclear Power Plants (NPPs). While plant procedures conventionally have been paper-based, computerized-based procedures are being implemented to reduce the drawbacks of paper-based procedures in many nuclear power plants. The Computerized Procedure System (CPS) designed by Korea Hydro and Nuclear Power Central Research Institute (KHNP CRI) is one of the human-system interfaces (HSIs) in digitalized Main Control Room (MCR) of APR1400 (Advanced Power Reactor 1400 MWe). Currently, CPS is being applied to constructing nuclear power plants of Korea and Barakah NPP 1, 2, 3 and 4 units of United Arab Emirates. The CPS has many advantages to perform the procedure in fully digitalized MCR. First, CPS provides the procedure flow with logic diagram to operators. The operator easily can be aware of the procedure flow from a previous instruction to the next instruction and also can find out the relation between parent instruction and child instructions such as AND, OR and SEQUENCE logics. Second, CPS has three logic-based functions such as procedure entry condition monitoring logic, continuously applied step (CAS) re-execution monitoring logic and auto evaluation logic on instructions. E.g. CPS provides the standard post trip actions procedure open popup message when the reactor trips by calculating the entry condition logic that procedure writer had made in the writing process. Third, CPS can directly display the task information related to instructions such as valves, pumps, process parameters, etc. and also the operator can call the system display related to procedure execution. If an operator clicks the system display link, the related system display popups on the right side monitor of CPS display. Lastly, CPS supports the synchronization of procedure among the operators. This synchronization function helps operators to succeed the goal of procedure and improve the situation

  7. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  8. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 592 MW(e) (nominal gross) electric power generating plant equipped with a Babcock and Wilcox Company (B and W) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  9. Ageing of fibre reinforced polymer composite selected as a bearing material for Rams of 540 MWe fuelling machine

    International Nuclear Information System (INIS)

    Limaye, P.K.; Soni, N.L.; Agrawal, R.G.

    2006-01-01

    Fibre-reinforced-polymer-composite material has been suggested as a bearing material to overcome tribological problems witnessed during the testing of Ram assembly of the 540 MWe fuelling machine at RTD. After successful trials at B-Ram the composite material has been adapted for B-RAM, C-Ram and RDB head at fuelling machines being tested at RTD, Hall 7 and at Tarapur. Laboratory evaluations were also carried out at Tribology Lab RTD to study effect of radiation on the composite. Paper deals with the various aspects of life prediction of this material in term of wear and radiation damage. (author)

  10. Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

    International Nuclear Information System (INIS)

    Mansouri, S.

    1985-06-01

    This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

  11. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 578 MW(e) (nominal gross) electric power generating plant equipped with a Foster Wheeler Energy Corporation (FWEC) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  12. Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

  13. Control rod studies for alternative fuel cycles in the GA 1160 MW(e) high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Neef, H. J.

    1975-06-15

    The control system, which is investigated in this paper for both the low enriched uranium high enriched uranium/thorium fuel cycles, has been developed to control the General Atomics (GA) thorium fuel cycle 1160 MW(e) reactor. It has been shown in this investigation that its effectiveness in the low enriched and subsequent thorium cycle switch-over reactor is equivalent to the effectiveness in the thorium cycle. The shutdown margin in the low enriched core is even higher compared to the thorium core, mainly due to the presence of Pa-233 in the thorium cycle. As long as the fuel cycle for the thorium cycle is not closed with the recycling of U-233, the low enriched cycle will offer an attractive alternative. It was found that the GA 1160 MW(e) control system has enough built-in control rod capacity to accommodate the low enriched uranium cycle and to perform a later switch-over to a thorium-based fuel cycle.

  14. Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts

    International Nuclear Information System (INIS)

    Miller, Laurence F.; Khuram Khan, M.; Williams, Wesley; Mynatt, F.R.

    2002-01-01

    This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model

  15. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 4: Supplementary engineering data

    Science.gov (United States)

    1981-01-01

    The reference conceptual design of the Magnetohydrodynamic Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD is summarized. Main elements of the design are identified and explained, and the rationale behind them is reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates, and identification of engineering issues that should be reexamined are also given. The latest (1980-1981) information from the MHD technology program are integrated with the elements of a conventional steam power electric generating plant. Supplementary Engineering Data (Issues, Background, Performance Assurance Plan, Design Details, System Design Descriptions and Related Drawings) is presented.

  16. Magnetohydrodynamics MHD Engineering Test Facility ETF 200 MWe power plant. Conceptual Design Engineering Report CDER. Volume 3: Costs and schedules

    Science.gov (United States)

    1981-01-01

    The estimated plant capital cost for a coal fired 200 MWE electric generating plant with open cycle magnetohydrodynamics is divided into principal accounts based on Federal Energy Regulatory Commision account structure. Each principal account is defined and its estimated cost subdivided into identifiable and major equipment systems. The cost data sources for compiling the estimates, cost parameters, allotments, assumptions, and contingencies, are discussed. Uncertainties associated with developing the costs are quantified to show the confidence level acquired. Guidelines established in preparing the estimated costs are included. Based on an overall milestone schedule related to conventional power plant scheduling experience and starting procurement of MHD components during the preliminary design phase there is a 6 1/2-year construction period. The duration of the project from start to commercial operation is 79 months. The engineering phase of the project is 4 1/2 years; the construction duration following the start of the man power block is 37 months.

  17. Treatment and processing of the effluents and wastes (other than fuel) produced by a 900 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Giraud

    1983-01-01

    Effluents produced by a 900 MWe power plant, are of three sorts: gaseous, liquid and solid. According to their nature, effluents are either released or stored for decaying before being released to the atmosphere. The non-contaminated reactor coolant effluents are purified (filtration, gas stripping) and treated by evaporation for reuse. Depending upon their radioactive level, liquid waste is either treated by evaporation or discharged after filtration. Solid waste issuing from previous treatments (concentrates, resins, filters) is processed in concrete drums using an encapsulation process. The concrete drum provides biological self-protection consistent with the national and international regulations pertaining to the transport of radioactive substance. Finally, the various low-level radioactive solid waste collected throughout the plant, is compacted into metal drums. Annual estimates of the quantity of effluents (gaseous, liquid) released in the environment and the number of drums (concrete, metal) produced by the plant figure in the conclusion

  18. The 900 MWe water pressurized reactor safety re-examination at the occasion of their third decennial inspection

    International Nuclear Information System (INIS)

    2009-01-01

    This document reports the safety re-examination actions performed on the French 900 MWe water pressurized reactors. This process includes three stages. The first one is an inventory of safety, design and operation requirements which are defined or specified in different texts: regulations, rules, criteria and specifications. This leads to compliance studies with respect to these documents and by in situ inspections, and then to corrective recommendations. After presenting this process, the report deals with specific safety studies which are related to external or internal aggressions (fire, explosions, flooding, climate, seism), to accidental situations (primary circuit cold overpressure, severe accidents, containment, level 1 and 2 safety probabilistic studies, passive failure of safeguard circuits, vapour generator tube failure, and so on), to design and sizing of civil engineering works and systems (radioactivity measurement system, safety injection system, recirculation function liability, liability of the irradiated fuel deactivation pool cooling system)

  19. Security of nuclear power in operation. Results from the first PWR 900 MWe stages of Electricity of France (EDF)

    Energy Technology Data Exchange (ETDEWEB)

    Capel, R; Chaubaron, J F [Electricite de France, 93 - Saint-Denis. Service de la Production Thermique

    1980-06-01

    The security and reliability objectives of the PWR 900 MWe stages are acquiring particular importance in the present energetic and nuclear context. This article presents the general framework wherein the superintendence and maintenance of plant equipment are situated. E.D.F. applies to all of its activities, the assurance of quality principles. The General Rules of Operating constitute the basic document. The Operating Technical Specifications specify the conditions for the correct operating and safety of the installations. The Organization of Quality handbook sets the rules to be obeyed in the management of all operations. Examples from Fessenhein and Bugey illustrate the subject and elucidate the practical dimension of security. Lastly, the lessons of experience are recalled.

  20. Some aspects of optimising the reactor core for a 600 MW(e) high temperature helium turbine power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Presser, W

    1972-04-24

    For the HHT 600 MW(e) power plant a core design with Triagonal blocks containing 24 channels with directly cooled fuel pins was considered. The design was found to require a low HM loading in the fuel zone to achieve favourable economic merits. For low HM densities a strong incentive exists to aim for burn-ups between 80 and 100 GWd/t. At the present an average discharge irradiation of 80 GWd/t was thought feasible and a reference design with a HM density of 0.6 g/cm {sup 3} in both core zones was chosen. The optimisation is not likely to be upset by local hot channel effects as a special investigation into the influence of safety margins found no changes in fuel cycle economics.

  1. Evaluation of ultimate load bearing capacity of the primary containment of a typical 540 MWe Indian PHWR

    International Nuclear Information System (INIS)

    Ray, Indrajit; Roy, Raghupati; Verma, U.S.P.; Warudkar, A.S.

    2003-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical 540 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) under beyond postulated design basis accident (DBA) scenario. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses about the study carried out for mesh sensitivity of the finite element (FE) discretization on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS under beyond postulated DBA scenario has been evaluated. (author)

  2. Seismic analysis of two 1050 mm diameter heavy water upgrading towers for 235 MWe Kaiga Atomic Power Plant Site

    International Nuclear Information System (INIS)

    Soni, R.S.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.; Narwaria, Suresh; Vardarajan, T.G.; Sadhukhan, H.K.

    1992-01-01

    This report deals with the analysis carried out for the evaluation of earthquake induced stresses and deflections in two 1050 mm diameter heavy water upgrading towers for Kaiga Atomic Power Plant Site. The analysis of upgrading tower has been carried out for two mutually perpendicular horizontal excitations and one vertical excitation applied simultaneously. The upgrading towers have been analysed using beam model taking into account soil-structure interaction. Response spectrum analysis has been carried out using site spectra for 235 MWe Kaiga reactor site. The seismic analysis has been performed for both the towers with supporting structure along with concrete pedestals and raft foundation. The towers have been checked for its stability due to compressive stresses to avoid buckling so that the nearby safety related structures are not geopardised in the event of safe shutdown earthquake (SSE) loading. (author). 14 refs., 12 figs., 18 tabs

  3. Model, parameter and code of environmental dispersion of gaseous effluent under normal operation from nuclear power plant with 600 MWe

    International Nuclear Information System (INIS)

    Hu Erbang; Gao Zhanrong

    1998-06-01

    The model of environmental dispersion of gaseous effluence under normal operation from a nuclear power plant with 600 MWe is established to give a mathematical expression of annual mean atmospheric dispersion factor under mixing release condition based on quality assessment of radiological environment for 30 years of Chinese nuclear industry. In calculation, the impact from calm and other following factors have been taken into account: mixing layer, dry and wet deposition, radioactive decay and buildings. The doses caused from the following exposure pathways are also given by this model: external exposure from immersion cloud and ground deposition, internal exposure due to inhalation and ingestion. The code is named as ROULEA. It contains four modules, i.e. INPUT, ANRTRI, CHIQV and DOSE for calculating 4-dimension joint frequency, annual mean atmospheric dispersion factor and doses

  4. Modeling and simulations of a 30 MWe solar electric generating system using parabolic trough collectors in Turkey

    Energy Technology Data Exchange (ETDEWEB)

    Usta, Yasemin [Anyl Asansor Ltd (Turkey)], email: syusta@gmail.com; Baker, Derek [Middle East Technical University (Turkey)], email: dbaker@metu.edu.tr; Kaftanoglu, Bilgin [Atilim University (Turkey)], email: bilgink@atilim.edu.tr

    2011-07-01

    With the energy crisis and the increasing concerns about climate change, the interest in concentrating solar power (CSP) systems is growing in Turkey. The aim of this paper is to develop a model of a CSP system using a field of parabolic trough collectors and to assess the predicted performance of the system. A model was developed for a 30MWe solar generating system in Antalya, Turkey, using TRNSYS software, the solar thermal electric components library and information on an existing system in Kramer Junction, California, United States. Annual simulations were then performed for both systems in Antalya and California using weather data. It was found that the predictions were in good agreement with published data. In addition results showed that Antalya's system would generate 30% less than Kramer Junction's system on an annual basis. This paper provides useful information on modeling and simulation of CSP systems.

  5. Study on the muon spectra at the depth of 570 m.w.e. underground with 100t scintillation detector

    International Nuclear Information System (INIS)

    Enikeev, R.I.; Zatsepin, G.T.; Korol'kova, E.V.; Kudryavtsev, V.A.; Mal'gin, A.S.; Ryazhskaya, O.G.; Khal'chugov, F.F.

    1988-01-01

    The experiment was carried out with 100-ton scintillation detector placed in the salt mine at the depth of 570 m.w.e. Detector measured the spectrum of energy release of electromagnetic cascades generated by muons underground. Electromagnetic and nuclear cascades were separated by the number of neutrons contained in the cascades. The measured spectrum of energy releases agrees with π- and K-meson spectrum with γ π,K =1.75±0.08 for muon energies at sea level E μ 0 > 0.7 TeV. The experimental data transformed to the vertical muon spectrum at sea level are in good agreement with the results of other works. The primary cosmic ray spectrum and the characteristics of pA-interactions up to energies of ∼ 100 TeV have not a changes which would lead to the increase of the γ π,K value higher than 1.85

  6. Living probabilistic safety assessment of French 1300 MWe PWR nuclear power plant unit: methodology, results and teaching

    International Nuclear Information System (INIS)

    Dubreuil Chambardel, A.; Villemeur, A.; Berger, J.P.; Moroni, J.M.

    1991-02-01

    Launched in 1986 by Electricite de France, the Probabilistic Safety Assessment of a French 1300 MWe Pressurized Water Reactor (called PSA 1300) was completed in 1989. The first objective was to assess the annual core damage frequency by identifying all the accident scenarii likely to contribute significantly to this frequency. The second objective of the study was to provide an automated computerized tool (software) for updating the assessment - in order to take new data and knowledge into account - and for performing numerous sensitivity studies easily. Its scope and characteristics render this study unique. Indeed, it required an effort amounting to 50 engineer-years. The results and the first lessons are presented in this paper. The PSA 1300 teachings will be extensively used for the design and operation of existing or future French nuclear power reactors

  7. Influence of declivitous secondary air on combustion characteristics of a down-fired 300-MWe utility boiler

    Energy Technology Data Exchange (ETDEWEB)

    Zhengqi Li; Feng Ren; Zhichao Chen; Zhao Chen; Jingjie Wang [Harbin Institute of Technology, Harbin (China). School of Energy Science and Engineering

    2010-02-15

    Industrial experiments were performed with a 300-MWe full-scale down-fired boiler. New data is reported for (i) gas temperature distributions within the primary air and coal mixture flows, (ii) gas compositions, such as O{sub 2}, CO, CO{sub 2} and NOx, and (iii) gas temperatures within the near-wall region. The data complements previously-obtained data from the same utility boiler before being modified by declination of the F-tier secondary air. By directing secondary air under the arches, the region where the primary air and pulverized coal mixture is ignited is brought forward within the boiler. Gas temperatures rose in the fuel-burning zone and fell in the fuel-burnout zone. As a result the quantity of unburned carbon in fly ash and the gas temperature at the furnace outlet were both lowered. 20 refs., 7 figs., 2 tabs.

  8. Effect of combustion characteristics on wall radiative heat flux in a 100 MWe oxy-coal combustion plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.; Ryu, C. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Chae, T.Y. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Yang, W. [Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Kim, Y.; Lee, S.; Seo, S. [Korea Electric Power Research Institute (KEPRI), Daejeon (Korea, Republic of). Power Generation Lab.

    2013-07-01

    Oxy-coal combustion exhibits different reaction, flow and heat transfer characteristics from air-coal combustion due to different properties of oxidizer and flue gas composition. This study investigated the wall radiative heat flux (WRHF) of air- and oxy-coal combustion in a simple hexahedral furnace and in a 100 MWe single-wall-fired boiler using computational modeling. The hexahedral furnace had similar operation conditions with the boiler, but the coal combustion was ignored by prescribing the gas properties after complete combustion at the inlet. The concentrations of O{sub 2} in the oxidizers ranging between 26 and 30% and different flue gas recirculation (FGR) methods were considered in the furnace. In the hexahedral furnace, the oxy-coal case with 28% of O{sub 2} and wet FGR had a similar value of T{sub af} with the air-coal combustion case, but its WRHF was 12% higher. The mixed FGR case with about 27% O{sub 2} in the oxidizer exhibited the WRHF similar to the air-coal case. During the actual combustion in the 100 MWe boiler using mixed FGR, the reduced volumetric flow rates in the oxy-coal cases lowered the swirl strength of the burners. This stretched the flames and moved the high temperature region farther to the downstream. Due to this reason, the case with 30% O{sub 2} in the oxidizers achieved a WRHF close to that of air-coal combustion, although its adiabatic flame temperature (T{sub af}) and WHRF predicted in the simplified hexahedral furnace was 103 K and 10% higher, respectively. Therefore, the combustion characteristics and temperature distribution significantly influences the WRHF, which should be assessed to determine the ideal operating conditions of oxy- coal combustion. The choice of the weighted sum of gray gases model (WSGGM) was not critical in the large coal-fired boiler.

  9. Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

    2009-01-01

    The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

  10. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series)

    International Nuclear Information System (INIS)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-01-01

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors)

  11. Evolution of new X and Y positioning system for 540 MWe PHWR fuelling machines - based on commissioning experience

    International Nuclear Information System (INIS)

    Gupta, Vivek; Vyas, A.K.; Gupta, K.S.; Rama Mohan, N.; Bhambra, H.S.

    2006-01-01

    In PHWR units, X and Y positioning system is provided to give feedback regarding the misalignment between end-fitting and Fuelling Machine (FM) Head during homing on process for carrying out the correction before clamping the Head. The existing design of X and Y Positioning System works by measuring the misalignment by sensing the tilt of the FM Head in X and Y direction caused by its mechanical interfacing with end-fitting as it is advanced in Z direction. The misalignment of Head is corrected by moving it in X and Y direction by X-fine and Y-fine drives, at Z pre-stop position. This correction is vital for achieving the satisfactory sealing of heavy water from channel at snout of FM Head with end fitting. During testing and commissioning trials, it was found that the end fitting of 540 MWe coolant channel assembly either tilts or bends due to the application of load by Fuelling Machines during the process of homing-on of FM Head. Due to this phenomenon, value of misalignment sensed by the Positioning System was considerably lower than the actual misalignment and consequently results in uncorrected misalignment. It was also observed that the high unbalanced moments caused by movement of heavier mass of B-ram in FM Head was further aggravating the misalignment problem. The problem, as an interim measure, was solved by optimising the loads acting on the end fitting to achieve the practically minimum possible uncorrected misalignment. However, to provide a lasting solution for this problem, a new X and Y Positioning System has been evolved. In this system, the misalignment between FM Head and end fitting is found by direct actuation of linear Variable Differential Transformer (LVDT) sensors by four separate alignment plates mounted on the snout. Further development to evolve a completely non-invasive technique using laser sensors has also been undertaken. This paper describes the problems encountered during commissioning of existing design of X and Y Positioning

  12. Methods and results of a PSA level 2 for a German BWR of the 900 MWe class

    International Nuclear Information System (INIS)

    Loffler, H.; Sonnenkalb, M.

    2006-01-01

    On behalf of the federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU) GRS has performed a PSA level 2 for a BWR type 69 NPP of the 900 MWe class, equipped with a N 2 inerted steel containment and a pressure suppression system. Integral deterministic accident analyses have been performed with the computer code MELCOR 1.8.5. Additional analyses have been done for those events and phenomena which are not or not sufficiently covered by MELCOR. The probabilistic event tree analysis begins with the core damage states received from PSA level 1, and it ends with the definition of release categories and the determination of their frequencies. Uncertainties about the frequency of core damage states and about events during the accident progression are taken into account by means of Monte Carlo simulations. If there is a core damage state there is a high probability (>50 %) for a very high and rapid release of radionuclides into the environment. This high conditional probability is due to the very low probability to retain a partly destroyed core inside the reactor pressure vessel (RPV) and because the containment almost certainly fails at the bottom of the control rod drives room after melt release from the failed RPV. (authors)

  13. Stress analysis of secondary ramp and secondary tilting mechanism of inclined fuel transfer machine for 500 MWe PFBR

    International Nuclear Information System (INIS)

    Prabhakaran, K.M.; Vaze, K.K.; Ghosh, A.K.; Rai, Somesh; Sundarani, A.R.; Patel, R.J.; Agrawal, R.G.

    2004-10-01

    Inclined Fuel Transfer Machine (IFTM) is one of the important machine of the fuel handling system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It is used to transfer core sub-assemblies (CSA) from reactor vessel to fuel building and vice-versa. Secondary ramp and Secondary tilting mechanism (SR/STM) is a part of IFTM which acts as a passage to transfer CSA. This mechanism and components were designed by the Refuelling Technology Division of BARC as per the ASME design code as class 2 component. Being critical in nature and complicated in geometry it was required to check the design of these components by detailed finite element analysis. The loading considered in the present study was static, thermal and seismic conditions. This was done using FEM software COSMOS/M. The Stresses were categorised as per the requirement of the ASME code for various levels of loading (Level A, B and C). Based on the analysis performed, it was concluded that the SR/STM qualifies the requirement of ASME code Section-III NC (Class-2 components). This report gives the details of the studies performed. (author)

  14. Elasto-plastic finite element analysis of axial surface crack in PHT piping of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Chawla, D.S.; Bhate, S.R.; Kushwaha, H.S.; Mahajan, S.C.

    1994-01-01

    The leak before break (LBB) approach in nuclear piping design envisages demonstrating that the pressurized pipe with a postulated flaw will leak at a detectable rate leading to corrective action well before catastrophic rupture would occur. This requires analysis of cracked pipe to study the crack growth and its stability. This report presents the behaviour of a surface crack in the wall of a thick primary heat transport (PHT) pipe of 500 MWe Indian PHWR. The line spring model (LSM) finite element is used to model the flawed pipe geometry. The variation of crack driving force (J-integral) across the crack front has been presented. The influence of crack geometry factors such as depth, shape, aspect ratio, and loading on peak values of J-integral as well as crack mouth opening displacement has been studied. Several crack shapes have been used to study the shape influence. The results are presented in dimensionless form so as to widen their applicability. The accuracy of the results is validated by comparison with results available in open literature. (author). 47 refs., 8 figs

  15. Comparison of safety margins for leak-before-break assessment of 500 MWe PHWR straight pipes: using contemporary techniques

    International Nuclear Information System (INIS)

    Rastogi, Rohit; Bhasin, Vivek; Kushwaha, H.S.

    1998-01-01

    The Leak Before Break (LBB) analysis of Primary Heat Transport (PHT) Piping of 500 MWe Indian PHWR is being performed using different well established techniques like R6 method (Nuclear Electric UK) and J-Tearing based methods (USNRC). These methods show that PHT piping has required safety margins and can be qualified for LBB. These analysis also showed that the piping has high fracture toughness and plastic collapse is the dominant mode of failure. To enhance the confidence in the results obtained from the above methods, further studies were done on the PHT piping. Procedures which predicted margins against plastic collapse were used. The analysis procedures used were Modified Limit Load Method, MPA Method (both from Germany), Moments Method (from Italy) and the Z-Factor method given in ASME Boiler and Pressure Vessel Code. The safety margins obtained from these analysis satisfied the LBB requirements. A table was generated which compared the safety margins obtained using all the above mentioned procedures. This report presents the results of this study. (author)

  16. Effect of Flow Configuration on Velocity and Temperature Distribution of Moderator Inside 540 MWe PHWR Calandria using CFD Techniques

    International Nuclear Information System (INIS)

    Bharj, J.S.; Sahaya, R.R.; Datta, D.; Dharne, S.P.

    2006-01-01

    The calandria of a Pressurized Heavy Water Reactor (PHWR) is a horizontal cylindrical vessel housing a matrix of horizontal tubes called calandria tubes, through which pass the pressure tubes that house the fuel bundles. The calandria is filled with heavy water acting as moderator. A large amount of heat (about 95 MW) is generated within the moderator mainly due to neutron slowing down and attenuation of gamma radiations. In the present configuration of 540 MWe calandria, moderator inlet diffusers are directed upwards and the outlet is from the bottom of the calandria. This configuration is not conducive for the buoyancy-dominated flows generated due to large volumetric heat generation in the moderator. In order to decide the effects of changes in flow configuration by changing location/direction of inlet/outlet nozzles, a study was done for moderator flows in the using PHOENICS CFD software. The results of study with various flow configurations show that modification in moderator flow configuration, reduces the peak temperature of moderator in calandria by about 12 deg C as well as gives a much more uniform temperature distribution. (authors)

  17. Test of small-scale central-core-cavity closure for a 300-MW(e) GCFR

    International Nuclear Information System (INIS)

    Robinson, G.C.; Dougan, J.R.; Naus, D.J.

    1981-01-01

    Under the Prestressed Concrete Reactor Vessel (PCRV) Program at the Oak Ridge National Laboratory, model tests are conducted to verify the design of the PCRV for a 300 MW(e) Gas-Cooled Fast Reactor (GCFR). Prominent features of the 1:20-scale central core cavity model included a close pitched array of fifty-five penetration tubes, forty-four segmented gusset/bearing plate assemblies, and intermeshed reinforcing steel. The closure model which was designed for a maximum cavity pressure (MCP) of 10.08 MPa was initially tested by applying 10 pressurization cycles from essentially no load to the MCP with strain and deflection data obtained during each cycle. This was followed by pressurization cycles to 32.8 MPa, 41.3 MPa, 48.3 MPa, 58.4 MPa and 79.3 MPa. At a pressure of 79.3 MPa an end cap on a penetration tube developed leaks and the test was terminated. An inelastic analysis was conducted to provide an estimate of the ultimate strength of the closure plug and to determine the potential mode of failure

  18. Design of a multivariable controller for a CANDU 600 MWe nuclear power plant using the INA method

    International Nuclear Information System (INIS)

    Roy, N.; Boisvert, J.; Mensah, S.

    1984-04-01

    The development of large and complex nuclear and process plants requires high-performance control systems, designed with rigorous multivariable techniques. This work is part of an analytical study demonstrating the real potential of multivariable methods. It covers every step in the design of a multi-variable controller for a Gentilly-2 type CANDU 600 MWe nuclear power plant using the Inverse Nyquist Array (INA) method. First the linear design model and its preliminary modifications are described. The design tools are reviewed and the operations required to achieve open-loop diagonal dominance are thoroughly described. Analysis of the closed-loop system is then performed and a feedback matrix is selected to meet the design specifications. The performance of the controller on the linear model is verified by simulation. Finally, the controller is implemented on the reference non-linear model to assess its overall performance. The results show that the INA method can be used successfully to design controllers for large and complex systems

  19. Mathematical modelling of straw combustion in a 38 MWe power plant furnace and effect of operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yao Bin Yang; Robert Newman; Vida Sharifi; Jim Swithenbank; John Ariss [Sheffield University, Sheffield (United Kingdom). Sheffield University Waste Incineration Centre (SUWIC), Department of Chemical and Process Engineering

    2007-01-15

    As one of the most easily accessible renewable energy resources, straw can be burned to provide electricity and heat to local communities. In this paper, mathematical modelling methods have been employed to simulate the operation of a 38 MWe straw-burning power plant to obtain detailed information on the flow and combustion characteristics in the furnace and to predict the effect on plant performance of variation in operating conditions. The predicted data are compared to measurements in terms of burning time, furnace temperature, flue gas emissions (including NOx), carbon content in the ash and overall combustion efficiency. It is concluded that straw burning on the grate is locally sub-stoichiometric and most of the NO is formed in the downstream combustion chamber and radiation shaft; auxiliary gas burners are responsible for the uneven distribution of temperature and gas flow at the furnace exit; and fuel moisture content is limited to below 25% to prevent excessive CO emission without compromising the plant performance. The current work greatly helps to understand the operating characteristics of large-scale straw-burning plants. 33 refs., 15 figs., 3 tabs.

  20. A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

    THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

  1. Structural mechanics research and development for main components of chinese 300 MWe PWR NPPs: from design to life management

    International Nuclear Information System (INIS)

    Yao Weida; Dou Yikang; Xie Yongcheng; He Yinbiao; Zhang Ming; Liang Xingyun

    2005-01-01

    Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

  2. The Influence of atmospheric conditions to probabilistic calculation of impact of radiology accident on PWR 1000 MWe

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Sri Kuntjoro

    2015-01-01

    The calculation of the radiological impact of the fission products releases due to potential accidents that may occur in the PWR (Pressurized Water Reactor) is required in a probabilistic. The atmospheric conditions greatly contribute to the dispersion of radionuclides in the environment, so that in this study will be analyzed the influence of atmospheric conditions on probabilistic calculation of the reactor accidents consequences. The objective of this study is to conduct an analysis of the influence of atmospheric conditions based on meteorological input data models on the radiological consequences of PWR 1000 MWe accidents. Simulations using PC-Cosyma code with probabilistic calculations mode, the meteorological data input executed cyclic and stratified, the meteorological input data are executed in the cyclic and stratified, and simulated in Muria Peninsula and Serang Coastal. Meteorological data were taken every hour for the duration of the year. The result showed that the cumulative frequency for the same input models for Serang coastal is higher than the Muria Peninsula. For the same site, cumulative frequency on cyclic input models is higher than stratified models. The cyclic models provide flexibility in determining the level of accuracy of calculations and do not require reference data compared to stratified models. The use of cyclic and stratified models involving large amounts of data and calculation repetition will improve the accuracy of statistical calculation values. (author)

  3. Evaluation of ultimate load bearing capacity of the primary containment of a typical first generation 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Singh, A.K.; Ray, I.; Roy, R.; Garg, R.P.; Verma, U.S.P.

    2005-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical first generation 220 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) beyond postulated design basis accident (DBA) scenario. The first generation ICS of Indian PHWRs are made of cylindrical wall capped with prestressed/reinforced cellular containment slab, which is connected monolithically to outer containment (OC) Wall to provide clamping effect during postulated DBA scenario. This paper discusses the simulation of construction sequence in analytical model, which is a very important aspect from point of view of capturing the residual stresses generated in the structure. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses the study of mesh sensitivity of the finite element (FE) discretisation on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS beyond postulated DBA scenario has been evaluated. (authors)

  4. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  5. Development of high pressure conductivity probe (HPCP) for secondary shut down system (SDS-2) of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Mohan, L.R.

    2003-09-01

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface. This interface moves towards the calandria because of molecular diffusion, temperature difference and physical disturbances in the moderator level. It is proposed to install two numbers of high pressure conductivity probes (HPCP) to monitor the interface movement as well as to provide the safe annunciation value for interface location. On actuation of the SDS-2 signal, high-pressure helium will inject the poison into the moderator to shutdown the reactor. During poison injection, these probes will experience high pressure of nearly 85 kg/sq.cm. Global market survey indicated that conductivity probes having built in temperature sensor are available for a maximum pressure rating of 35 kg/sq.cm. Hence in order to meet the process requirement of SDS-2, the development of HPCP suitable for a pressure of 85 kg/sq.cm. was taken up. Two numbers of such probes were successfully designed, fabricated and evaluated for their performance. The developed conductivity probes fully meet the laid design and performance criteria. The aforesaid development work was a successful endeavour towards indigenisation of high-pressure conductivity probe for future applications. This report deals with the design aspects, fabrication technique, material and performance evajuation criteria and test results of HPCP. (author)

  6. Foster Wheeler's Solutions for Large Scale CFB Boiler Technology: Features and Operational Performance of Łagisza 460 MWe CFB Boiler

    Science.gov (United States)

    Hotta, Arto

    During recent years, once-through supercritical (OTSC) CFB technology has been developed, enabling the CFB technology to proceed to medium-scale (500 MWe) utility projects such as Łagisza Power Plant in Poland owned by Poludniowy Koncern Energetyczny SA. (PKE), with net efficiency nearly 44%. Łagisza power plant is currently under commissioning and has reached full load operation in March 2009. The initial operation shows very good performance and confirms, that the CFB process has no problems with the scaling up to this size. Also the once-through steam cycle utilizing Siemens' vertical tube Benson technology has performed as predicted in the CFB process. Foster Wheeler has developed the CFB design further up to 800 MWe with net efficiency of ≥45%.

  7. The effect of core design changes on the doubling time and the fuel cycle cost of a 1,000 MWe LMFBR

    International Nuclear Information System (INIS)

    Otake, I.; Inoue, T.; Tomabechi, K.; Osada, H.; Aoki, K.

    1978-01-01

    Core design studies were performed to improve the doubling time and to minimize the fuel cycle cost of a 1,000 MWe Fast Demonstration Reactor. A core was designed mainly based on the technology being used for the design of a prototype fast reactor MONJU, because much valuable experience will be forthcoming from this reactor. Design parameters with a wide variable range were used to clarify the relations between breeding characteristics, fuel economics and various designs. (author)

  8. Severe accident mitigation strategy for the generation II PWRs in France. Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

    Energy Technology Data Exchange (ETDEWEB)

    Cenerino, G.; Rahni, N.; Chevrier, P.; Dubreuil, M.; Guigueno, Y.; Raimond, E.; Bonnet, J.M. [IRSN/PSN-RES/SAG (France)

    2013-07-01

    The 3{sup rd} Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured. Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications. Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define. Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others. (orig.)

  9. Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

    International Nuclear Information System (INIS)

    Chang, Si Young; Ha, Chung Woo

    1988-01-01

    An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

  10. Process simulation of co-firing torrefied biomass in a 220 MWe coal-fired power plant

    International Nuclear Information System (INIS)

    Li, Jun; Zhang, Xiaolei; Pawlak-Kruczek, Halina; Yang, Weihong; Kruczek, Pawel; Blasiak, Wlodzimierz

    2014-01-01

    Highlights: • The performances of torrefaction based co-firing power plant are simulated by using Aspen Plus. • Mass loss properties and released gaseous components have been studied during biomass torrefaction processes. • Mole fractions of CO 2 and CO account for 69–91% and 4–27% in total torrefied gases. • The electrical efficiency reduced when increasing either torrefaction temperature or substitution ratio of biomass. - Abstract: Torrefaction based co-firing in a pulverized coal boiler has been proposed for large percentage of biomass co-firing. A 220 MWe pulverized coal-power plant is simulated using Aspen Plus for full understanding the impacts of an additional torrefaction unit on the efficiency of the whole power plant, the studied process includes biomass drying, biomass torrefaction, mill systems, biomass/coal devolatilization and combustion, heat exchanges and power generation. Palm kernel shells (PKS) were torrefied at same residence time but 4 different temperatures, to prepare 4 torrefied biomasses with different degrees of torrefaction. During biomass torrefaction processes, the mass loss properties and released gaseous components have been studied. In addition, process simulations at varying torrefaction degrees and biomass co-firing ratios have been carried out to understand the properties of CO 2 emission and electricity efficiency in the studied torrefaction based co-firing power plant. According to the experimental results, the mole fractions of CO 2 and CO account for 69–91% and 4–27% in torrefied gases. The predicted results also showed that the electrical efficiency reduced when increasing either torrefaction temperature or substitution ratio of biomass. A deep torrefaction may not be recommended, because the power saved from biomass grinding is less than the heat consumed by the extra torrefaction process, depending on the heat sources

  11. Design and development of improved ballscrew and control circuit for reactivity mechanisms of 220 MWe PHWR operating stations

    International Nuclear Information System (INIS)

    Jain, A.K.; Rama Mohan, N.; Mathew, Jimmy; Mathur, M.K.; Roy, S.; Ingle, V.J.; Ghoshal, B.; Ashok Kumar, B.; Patil, D.C.; Dwivedi, K.P.; Bhambra, H.S.

    2006-01-01

    There has been persistent failure of Ballscrews used for Reactivity Mechanism in standardised 220 MWe PHWR units. The detailed review of failures indicated that on one hand the number of demands for operation of Absorber Rod and Regulating Rod had increased due to use of digital circuit in the drive control system as compared analog circuits used earlier. On the other hand, the existing design of ballscrew had some inherent weaknesses to withstand the loads generated during starting and stopping of the regulating rods. To solve these problems two-pronged approach was adopted. The control problem was traced to overshooting of the servomotor of Absorber Rod and Regulating Rod to the full speed at the time of starting and thereafter, settling to the required speed. This sudden overshooting produces a jerk in the drive mechanism. A modified circuit has been evolved to solve this problem. Also, Changing the dead band and gain of control circuits have reduced the number of rod movements. A 'new design' of Ballscrew assembly was finalised by NPCIL with a view to withstand the severe loads generated during starting and stopping of the regulating rods and to achieve enhanced service life under water-lubrication condition. Based on this design, prototype assemblies were successfully manufactured by two Indian manufacturers. The design was cleared for manufacturing of the bulk production of Ballscrew assemblies after evaluation of its performance during rigorous 'Acceptance Testing'. Two ballscrews of new design were installed in the KGS-1 reactor and are operating since July 2005. This paper covers operational feedback including ballscrew failures in various units, Design/Development of Modified Reactivity Mechanism Ballscrews and Control Circuit based on analysis of underlying causes of failures and feedback on performance of new design. (author)

  12. Thermoeconomic Modeling and Parametric Study of a Photovoltaic-Assisted 1 MWe Combined Cooling, Heating, and Power System

    Directory of Open Access Journals (Sweden)

    Alexandros Arsalis

    2016-08-01

    Full Text Available In this study a small-scale, completely autonomous combined cooling, heating, and power (CCHP system is coupled to a photovoltaic (PV subsystem, to investigate the possibility of reducing fuel consumption. The CCHP system generates electrical energy with the use of a simple gas turbine cycle, with a rated nominal power output of 1 MWe. The nominal power output of the PV subsystem is examined in a parametric study, ranging from 0 to 600 kWe, to investigate which configuration results in a minimum lifecycle cost (LCC for a system lifetime of 20 years of service. The load profile considered is applied for a complex of households in Nicosia, Cyprus. The solar data for the PV subsystem are taken on an hourly basis for a whole year. The results suggest that apart from economic benefits, the proposed system also results in high efficiency and reduced CO2 emissions. The parametric study shows that the optimum PV capacity is 300 kWe. The minimum lifecycle cost for the PV-assisted CCHP system is found to be 3.509 million €, as compared to 3.577 million € for a system without a PV subsystem. The total cost for the PV subsystem is 547,445 €, while the total cost for operating the system (fuel is 731,814 € (compared to 952,201 € for a CCHP system without PVs. Overall the proposed system generates a total energy output of 210,520 kWh (during its whole lifetime, which translates to a unit cost of 17 €/kWh.

  13. Burnup effects on criticality, breeding and safety of 1,000 MWe gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohta, Fumio

    1977-12-01

    Burnup characteristics of 1,000 MWe, PuO 2 - UO 2 fuelled helium-cooled fast breeder reactor have been studied concerning criticality, breeding and safety. A 26-energy group cross-section set produced from ENDF/B-3 was used. Criticality and breeding were studied with two-dimensional burnup code APOLLO and 4-energy group cross-section set generated by collapsing the mentioned cross-section set. Safety aspects such as Doppler reactivity effect, coolant-depressurisation and steam-ingression reactivity effect were studied with multi-dimensional diffusion theory code CITATION and perturbation theory code PERKY, as well as the 26-energy group cross-section set. The following were revealed: (1) The reactivity swing over a year's irradiation is merely 1.5% ΔK/K. This small swing may permit relatively long fuel dwelling in GCFR and , thus, the frequency of outages for refuelling can be minimised. (2) The surplus fissile plutonium over a year's irradiation is about 360 Kg, and the system doubling time is about 9 years. The GCFR studied has excellent breeding, compared with those in PuO 2 -UO 2 fuelled LMFBR and other GCFRs. (3) The coolant-depressurisation reactivity effect becomes more positive with burnup. This is not so serious as the sodium-void reactivity effect of LMFBR. (4) In the start-up core, the steam-ingression reactivity effect due to steam ingression to the core and blanket from the secondary coolant system becomes positive at certain steam density (0.02gr/cc) and this positive effect increases with steam density. With advance of burnup, however, the effect becomes negative, this increasing with steam density. After all, the steam ingression is no hazard in operation of GCFR since the reactivity effect is negative in the equilibrium state. (auth.)

  14. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Zheng Yanhua; Shi Lei; Wang Yan

    2010-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

  15. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts

    International Nuclear Information System (INIS)

    Bore, C.

    1995-01-01

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a 'viscosity pump' phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. In addition, final metallurgical

  16. Experimental study of poison moderator interface movement for shut down system #2(SDS#2) of 540 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Chawan, D.B.; Ananthan, P.; Sharma, B.S.V.G.; Mohan, L.R.

    2005-03-01

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface, termed as poison moderator interface (PMI). During normal operation of the reactor, the interface moves towards the calandria, mainly because of molecular diffusion from poison to moderator. Other reasons for movement are mixing of poison and moderator due to physical disturbances in the moderator level and to some extent due to temperature difference between the two liquids. The electrical conductivity of these liquids was found to be the most reliable parameter indicating interface movement. For this purpose, two on-line high-pressure conductivity probes have been installed on moderator side for each one of the six poison tanks. During normal operation of reactor, the interface moves slowly towards the calandria over a period of time and gives rise to increase in conductivity. To study the interface pattern and factors affecting the same, a full-scale experimental setup was developed and series of experiments carried out. The experimental results showed that the interface is quite stable and annunciation can be placed around 100 micro siemens/cm before back flushing is initiated. One dimensional diffusion analysis of the obtained experimental data showed that the derived model for PMI setup with diffusion parameter of 900 cm 2 /hr is able to predict the interface movement quite satisfactorily. This report gives an insight into the experiments carried out for estimation of the effective diffusion parameter for the poison moderator interface, model formulation and its prognostic behavior. (author)

  17. Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.N.; Mohan, L.R.

    2002-01-01

    Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

  18. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  19. Evolution of on-power refuelling system for 500 MWe PHWR based on experience from Rajasthan, Madras and Narora Atomic Power Stations

    International Nuclear Information System (INIS)

    Warrier, S.R.; Inder Jit; Sanatkumar, A.

    1991-01-01

    The on-power fuel handling system design at Rajasthan and Madras Atomic Power Stations (RAPS and MAPS) is essentially based on the design of the fuel handling system at Douglas Point Station (CANADA) Although, a number of improvements have been carried out in the fuel handling system of RAPS and MAPS at the component and sub-assembly level, some problems of repetitive nature like frequent deterioration in the performance of B-ram ball screw, leak detector solenoid valves etc., still exist. Further, there are certain limitations and drawbacks in the fuelling systems of these stations. For example, FM carriage design would not meet current seismic qualification standards. Also there are chances of fuel transfer room getting contaminated during movement of a failed fuel bundle. In order to obviate these deficiencies, a new concept has been worked out for the fuel handling system of Narora Atomic Power Station (NAPS) and accordingly, major changes have been made adopting a new layout. For example, FM head supporting arrangement has been changed to 'Suspension' type and a 'Linear-indexed' transfer magazine has been introduced in the fuel transfer system. Based on the experience gained from RAPS, MAPS and NAPS, design concept for 500 MWe fuel handling system has been evolved with further improvements especially in the layout. Also, a Calibration and Maintenance Facility for maintenance, testing calibration of FM head, sub-assemblies and components of fuel handling system has been introduced in the 500 MWe design. This paper discusses some of the experience gained from RAPS, MAPS and NAPS and also highlights the features of 500 MWe fuel handling system. (author)

  20. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 2: Engineering. Volume 3: Costs and schedules

    Science.gov (United States)

    1981-01-01

    Engineering design details for the principal systems, system operating modes, site facilities, and structures of an engineering test facility (ETF) of a 200 MWE power plant are presented. The ETF resembles a coal-fired steam power plant in many ways. It is analogous to a conventional plant which has had the coal combustor replaced with the MHD power train. Most of the ETF components are conventional. They can, however, be sized or configured differently or perform additional functions from those in a conventional coal power plant. The boiler not only generates steam, but also performs the functions of heating the MHD oxidant, recovering seed, and controlling emissions.

  1. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 2: Engineering. Volume 3: Costs and schedules. Final Report

    International Nuclear Information System (INIS)

    1981-09-01

    Engineering design details for the principal systems, system operating modes, site facilities, and structures of an engineering test facility (ETF) of a 200 MWE power plant are presented. The ETF resembles a coal-fired steam power plant in many ways. It is analogous to a conventional plant which has had the coal combustor replaced with the MHD power train. Most of the ETF components are conventional. They can, however, be sized or configured differently or perform additional functions from those in a conventional coal power plant. The boiler not only generates steam, but also performs the functions of heating the MHD oxidant, recovering seed, and controlling emissions

  2. Stopped cosmic-ray muons in plastic scintillators on the surface and at the depth of 25 m.w.e

    International Nuclear Information System (INIS)

    Maletić, D; Dragić, A; Banjanac, R; Joković, D; Veselinović, N; Udovicić, V; Savić, M; Anicin, I; Puzović, J

    2013-01-01

    Cosmic ray muons stopped in 5 cm thick plastic scintillators at surface and at depth of 25 m.w.e are studied. Apart from the stopped muon rate we measured the spectrum of muon decay electrons and the degree of polarization of stopped muons. Preliminary results for the Michel parameter yield values lower than the currently accepted one, while the asymmetry between the numbers of decay positrons registered in the upper and lower hemispheres appear higher than expected on the basis of numerous earlier studies.

  3. Development of process route for production of tubing for various core sub-assemblies and heat exchangers for 500 MWe Indian PFBR

    International Nuclear Information System (INIS)

    Lakshminarayana, B.; Phani Babu, C.; Dubey, A.K.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

    2009-01-01

    India's three stage Nuclear Power Program has entered its second stage on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies and control elements for this reactor. The challenging task of process development and production of the various critical tubing for these sub assemblies for PFBR has been taken up by Stainless Steel Tubes Plant (SSTP), NFC with indigenous development of the equipment and technology

  4. Estimated radiological effects of the normal discharge of radioactivity from nuclear power plants in the Netherlands with a total capacity of 3500 MWe

    International Nuclear Information System (INIS)

    Lugt, G. van der; Wijker, H.; Kema, N.V.

    1977-01-01

    In the Netherlands discussions are going on about the installation of three nuclear power plants, leading with the two existing plants to a total capacity of 3500 MWe. To have an impression of the radiological impact of this program, calculations were carried out concerning the population doses due to the discharge of radioactivity from the plants during normal operation. The discharge via the ventilation stack gives doses due to noble gases, halogens and particulate material. The population dose due to the halogens in the grass-milk-man chain is estimated using the real distribution of grass-land around the reactor sites. It could be concluded that the population dose due to the contamination of crops and fruit is negligeable. A conservative estimation is made for the dose due to the discharge of tritium. The population dose due to the discharge in the cooling water is calculated using the following pathways: drinking water; consumption of fish; consumption of meat from animals fed with fish products. The individual doses caused by the normal discharge of a 1000 MWe plant appeared to be very low, mostly below 1 mrem/year. The population dose is in the order of some tens manrems. The total dose of the 5 nuclear power plants to the dutch population is not more than 70 manrem. Using a linear dose-effect relationship the health effects to the population are estimated and compared with the normal frequency

  5. Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2010-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.

  6. Process Control Logic Modification to Mitigate Transient Following Tripping of a Primary Circulating Pump for a 540 MWe PHWR Power Plant

    International Nuclear Information System (INIS)

    Contractor, Ankur D; Gaikwad, Avinash J.; Kumar, Rajesh; Chakraborty, G.; Vhora, S.F.

    2006-01-01

    The 540 MWe Indian Pressurised Heavy Water Reactor (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like Primary Heat Transport (PHT) system configuration with two loops, four Primary Circulating Pumps (PCPs) and four passes through core, addition of a Pressurizer (surge Tank) in the PHT system along with Feed/Bleed system and their safety related implications, simulation model have been developed. A reactor step-back is proposed following one PCP trip. The corresponding PCP in the healthy loop is tripped to avoid asymmetrical flow and pressure distribution in the two identical loops. In spite of such elaborate provisions, the margins from high/low PHT pressure are small following tripping of one PCP. Mathematical models for all the major components and sub-systems of the proposed 540 MWe PHWR were developed based on the conservation equations of mass, momentum, energy and equation of state. All the associated control systems are also modeled. The PHT system includes the reactor core with nuclear fuel, PCP, PHT system pressure controller with feed/bleed system and Pressurizer (Surge Tank). The secondary system includes mainly the Steam Generators (SGs), the SG level and pressure controllers, apart from the various steam cycle components. All these models are integrated together to form the Plant Transient Analysis Computer Code Dyna540. The scenario following one PCP trips leads to different states (high/low pressure in Reactor Outlet Header (ROH)) depending upon the banks in which the PCP trips. The pressurizer is connected to two ROHs on one side of the reactor. The system pressure is controlled based on average of four ROHs pressure. In the case of asymmetrical pump operation, this logic leads to a situation where individual ROH pressure goes very near the low/high PHT system pressure trip set point, even though the controlled average pressure is very close to the set pressure. The PHT high

  7. CFD analysis of the pulverized coal combustion processes in a 160 MWe tangentially-fired-boiler of a thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Cristiano V. da; Beskow, Arthur B. [Universidade Regional Integrada do Alto Uruguai e das Misses (LABSIM/GEAPI/URI), Erechim, RS (Brazil). Dept. de Engenharia e Ciencia da Computacao. Grupo de Engenharia Aplicada a Processos Industriais], Emails: cristiano@uricer.edu.br, Arthur@uricer.edu.br; Indrusiak, Maria Luiza S. [Universidade do Vale do Rio dos Sinos (UNISINOS), Sao Leopoldo, RS (Brazil). Programa de Engenharia Mecanica], E-mail: sperbindrusiak@via-rs.net

    2010-10-15

    The strategic role of energy and the current concern with greenhouse effects, energetic and exegetic efficiency of fossil fuel combustion greatly enhance the importance of the studies of complex physical and chemical processes occurring inside boilers of thermal power plants. The state of the art in computational fluid dynamics and the availability of commercial codes encourage numeric studies of the combustion processes. In the present work the commercial software CFX Ansys Europe Ltd. was used to study the combustion of coal in a 160 MWe commercial thermal power plant with the objective of simulating the operational conditions and identifying factors of inefficiency. The behavior of the flow of air and pulverized coal through the burners was analyzed, and the three-dimensional flue gas flow through the combustion chamber and heat exchangers was reproduced in the numeric simulation. (author)

  8. Electronuclear reactors - EDF - Orientations of generic studies to be performed for the safety re-examination of 1300 MWe reactors associated to their third decennial inspection

    International Nuclear Information System (INIS)

    2011-01-01

    This report expresses the ASN's opinion on the framework and objectives of the EDF program concerning the generic studies of the safety re-examination of the 1300 MWe reactors associated with their third decennial inspection. This comprises lessons from the Fukushima accident, the improvement of the 'internal explosion' referential by using a probabilistic study, the application of the seismic margin assessment approach as soon as possible, checking the absence of any 'cliff effect' for cooling functions, a deepened re-examination of hurricane frequencies in France. Other request by the ASN concern the verification of the pertinence of release authorizations, taking the TSN law into account, taking the AP1300 project into account, the expansion of the complementary domain, the project of reactor lifetime extension. Some technical requests are discussed in appendix

  9. Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

    1996-01-01

    The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

  10. Running-in strategies for the low-enriched 600 MW(e) D-HHT reactor. Part 1. Comparison of different on-load refuelling schemes

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-14

    This paper presents detailed burn-up calculations and fuel management strategies for the Dragon-HHT, D-HHT, reference core. The reference layout was chosen from the outcome of a design survey with the 1-D equilibrium fuel cycle code FLATTER. The decision was based on aspects of engineering and economics. The purpose of the investigation is to devise a suitable first core, follow the irradiation history of the fuel and the general behaviour of the reactor during the first core replacements until equilibrium operating conditions are reached. A detailed description of time dependant burn-up and spatial power production for specified reactivity limits is required. For this purpose the reactor code system VSOP was employed. Different combinations of the parameters are investigated and the influence on reactor operation and economics discussed. From the strategy analysis a reference fuel management scheme is chosen for the low enriched 600 MW(e) D-HHT reactor.

  11. Opinion of the IRSN on serviceability of the 900 MWe reactor vessel - Answers to demands of the Nuclear Permanent Department of December 2005 - Mechanical aspect

    International Nuclear Information System (INIS)

    2010-05-01

    As three demands had been made to EDF in December 2005 regarding the serviceability and more particularly the mechanical behaviour of the 900 MWe reactor vessels, this report discusses the evolution brought to models and proposed by EDF to correct the defect plasticity and take residual stresses into account. This discussion notably concerns the defect height and length range, and the admissible residual stress, but also the use of safety coefficients, transient application, fluence and the brittle-ductile transition temperature. This report from the French Nuclear Safety and Radioprotection Institute (IRSN) outlines the failure risks associated to the vessel in some specific nuclear power stations. Recommendations are made regarding the residual stress amplitude, the risk of fracture by cleavage, and actions to correct fracture risk margins on vessels which do not comply with regulatory criteria

  12. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    A. Rama Rao

    2008-04-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  13. Experimental study of the tritium inventory in the BR3 and extrapolation to a P.W.R. of 900 MWe

    International Nuclear Information System (INIS)

    Charlier, A.; Gubel, P.; Vandenberg, C.; Haas, D.

    1982-01-01

    The aim of this report is to evaluate the tritium production and diffusion in uranium and plutonium fuel in the primary circuit of a PWR and to improve the knowledge about the production difference between the two kinds of isotopes. The first part of the work is relative to the experimental PWR BR3, cycle 4A, during which a constant control of the tritium activity has been performed in the primary circuit. These experimental evaluation was compared with the the theoretical estimation of the tritium production during the cycle 4A. From these observations and calculations, a tritium release fraction was deduced and estimated to be 0.81% of the total tritium produced in the fuel. The second part of the work is devoted to post-irradiation examinations on a few uranium and plutonium rods irradiated in the BR3 reactor. The tritium content was measured in the cladding, in the fuel and in the gas plenum for various samples of fuel rods. These results show the relationship between the release rate from the fuel matrix, the linear power and the burnup. The last part of the work is the estimate of the tritium production in a PWR of 900 MWe in operating conditions. The tritium production was calculated for an uranium fuelled core and for a core containing 30% of all plutonium fuel assemblies in a generic power plant of 900 MWe. From this study, it results that the loading with 30% plutonium assemblies at equilibrium increases the tritium balance in the moderator water of less than 5%

  14. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.2. Three-dimensional analysis of the temperature and stress fields in a HHT vessel, including effects of the thermal creep

    International Nuclear Information System (INIS)

    Rodriguez, C.; Rebora, B.

    1979-01-01

    The thermal rheological calculation of the prestressed concrete reactor vessel for the HHT-670 MW(e) Demonstration Plant is presented in the paper. The main aim of this calculation is to evaluate the effects of the elevated temperature and various loads on the liner as well as on the hot concrete

  15. Final Techno-Economic Analysis of 550 MWe Supercritical PC Power Plant CO2 Capture with Linde-BASF Advanced PCC Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, Devin [Linde LLC, Murray Hill, NJ (United States); Stoffregen, Torsten [Linde AG Linde Engineering Division, Dresden (Germany); Rigby, Sean [BASF Corporation, Houston, TX (United States)

    2017-01-09

    This topical report presents the techno-economic evaluation of a 550 MWe supercritical pulverized coal (PC) power plant utilizing Illinois No. 6 coal as fuel, integrated with 1) a previously presented (for a subcritical PC plant) Linde-BASF post-combustion CO2 capture (PCC) plant incorporating BASF’s OASE® blue aqueous amine-based solvent (LB1) [Ref. 6] and 2) a new Linde-BASF PCC plant incorporating the same BASF OASE® blue solvent that features an advanced stripper interstage heater design (SIH) to optimize heat recovery in the PCC process. The process simulation and modeling for this report is performed using Aspen Plus V8.8. Technical information from the PCC plant is determined using BASF’s proprietary thermodynamic and process simulation models. The simulations developed and resulting cost estimates are first validated by reproducing the results of DOE/NETL Case 12 representing a 550 MWe supercritical PC-fired power plant with PCC incorporating a monoethanolamine (MEA) solvent as used in the DOE/NETL Case 12 reference [Ref. 2]. The results of the techno-economic assessment are shown comparing two specific options utilizing the BASF OASE® blue solvent technology (LB1 and SIH) to the DOE/NETL Case 12 reference. The results are shown comparing the energy demand for PCC, the incremental fuel requirement, and the net higher heating value (HHV) efficiency of the PC power plant integrated with the PCC plant. A comparison of the capital costs for each PCC plant configuration corresponding to a net 550 MWe power generation is also presented. Lastly, a cost of electricity (COE) and cost of CO2 captured assessment is shown illustrating the substantial cost reductions achieved with the Linde-BASF PCC plant utilizing the advanced SIH configuration in combination with BASF’s OASE® blue solvent technology as compared to the DOE/NETL Case 12 reference. The key factors contributing to the reduction of COE and the cost of CO2 captured

  16. Influence of the overfire air ratio on the NO(x) emission and combustion characteristics of a down-fired 300-MW(e) utility boiler.

    Science.gov (United States)

    Ren, Feng; Li, Zhengqi; Chen, Zhichao; Fan, Subo; Liu, Guangkui

    2010-08-15

    Down-fired boilers used to burn low-volatile coals have high NO(x) emissions. To find a way of solving this problem, an overfire air (OFA) system was introduced on a 300 MW(e) down-fired boiler. Full-scale experiments were performed on this retrofitted boiler to explore the influence of the OFA ratio (the mass flux ratio of OFA to the total combustion air) on the combustion and NO(x) emission characteristics in the furnace. Measurements were taken of gas temperature distributions along the primary air and coal mixture flows, average gas temperatures along the furnace height, concentrations of gases such as O(2), CO, and NO(x) in the near-wall region and carbon content in the fly ash. Data were compared for five different OFA ratios. The results show that as the OFA ratio increases from 12% to 35%, the NO(x) emission decreases from 1308 to 966 mg/Nm(3) (at 6% O(2) dry) and the carbon content in the fly ash increases from 6.53% to 15.86%. Considering both the environmental and economic effect, 25% was chosen as the optimized OFA ratio.

  17. Improving combustion characteristics and NO(x) emissions of a down-fired 350 MW(e) utility boiler with multiple injection and multiple staging.

    Science.gov (United States)

    Kuang, Min; Li, Zhengqi; Xu, Shantian; Zhu, Qunyi

    2011-04-15

    Within a Mitsui Babcock Energy Limited down-fired pulverized-coal 350 MW(e) utility boiler, in situ experiments were performed, with measurements taken of gas temperatures in the burner and near the right-wall regions, and of gas concentrations (O(2) and NO) from the near-wall region. Large combustion differences between zones near the front and rear walls and particularly high NO(x) emissions were found in the boiler. With focus on minimizing these problems, a new technology based on multiple-injection and multiple-staging has been developed. Combustion improvements and NO(x) reductions were validated by investigating three aspects. First, numerical simulations of the pulverized-coal combustion process and NO(x) emissions were compared in both the original and new technologies. Good agreement was found between simulations and in situ measurements with the original technology. Second, with the new technology, gas temperature and concentration distributions were found to be symmetric near the front and rear walls. A relatively low-temperature and high-oxygen-concentration zone formed in the near-wall region that helps mitigate slagging in the lower furnace. Third, NO(x) emissions were found to have decreased by as much as 50%, yielding a slight decrease in the levels of unburnt carbon in the fly ash.

  18. Reducing NOx Emissions for a 600 MWe Down-Fired Pulverized-Coal Utility Boiler by Applying a Novel Combustion System.

    Science.gov (United States)

    Ma, Lun; Fang, Qingyan; Lv, Dangzhen; Zhang, Cheng; Chen, Yiping; Chen, Gang; Duan, Xuenong; Wang, Xihuan

    2015-11-03

    A novel combustion system was applied to a 600 MWe Foster Wheeler (FW) down-fired pulverized-coal utility boiler to solve high NOx emissions, without causing an obvious increase in the carbon content of fly ash. The unit included moving fuel-lean nozzles from the arches to the front/rear walls and rearranging staged air as well as introducing separated overfire air (SOFA). Numerical simulations were carried out under the original and novel combustion systems to evaluate the performance of combustion and NOx emissions in the furnace. The simulated results were found to be in good agreement with the in situ measurements. The novel combustion system enlarged the recirculation zones below the arches, thereby strengthening the combustion stability considerably. The coal/air downward penetration depth was markedly extended, and the pulverized-coal travel path in the lower furnace significantly increased, which contributed to the burnout degree. The introduction of SOFA resulted in a low-oxygen and strong-reducing atmosphere in the lower furnace region to reduce NOx emissions evidently. The industrial measurements showed that NOx emissions at full load decreased significantly by 50%, from 1501 mg/m3 (O2 at 6%) to 751 mg/m3 (O2 at 6%). The carbon content in the fly ash increased only slightly, from 4.13 to 4.30%.

  19. Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

    1992-01-01

    This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

  20. Comparison of control rod effectiveness for thorium and low-enriched fuel cycles in the GA-1, 160 MW(e) design

    Energy Technology Data Exchange (ETDEWEB)

    Neef, Hans Joachim

    1974-03-15

    In an investigation of the properties of the Thorium-Uranium (Th) and the Low-Enriched Uranium (LEU) fuel cycles it is also necessary to compare the effectiveness of the control rods in a reactor system operating with these sorts of fuel. Furthermore, it is under consideration to start a reactor with LEU fuel and switch-over to a Th cycle. It is also of interest to look at the switch-over phase in respect to the control rod effectiveness. The various fuel cycles have been studied for the same fuel element and control rod design, namely the one of GA's commercially available 1,160 MW(e) reference power station. This paper gives the first results on the control rod calculations and is presented mainly in two parts. Part 1 describes spectral effects which have been investigated by cell calculations with a discrete ordinates transport code. The main result is the higher effectiveness of a rod in a Th-cycle compared with a LEU-cycle. Part 2 reports on reactor calculations with a diffusion code and shows that this advantage can partially disappear in the reactor because of the spatial flux distribution. This effect has to be studied in further investigations for a full understanding.

  1. Conceptual design considerations for providing hook-up type schemes for tracking beyond design basis events (BDBE) for 700 MWe PHWR project

    International Nuclear Information System (INIS)

    Vhora, S.F.; Inder Jit; Bhardwaj, S.A.

    2005-01-01

    A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

  2. Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Singh, Ashutosh Pratap; Rajagopalan, C.; Rakesh, V.; Rajendran, S.; Venugopal, S.; Kasiviswanathan, K.V.; Jayakumar, T.

    2011-01-01

    Highlights: → Conceptual study on the configuration of an ISI device for FBR interspace environment has been carried out. → Prototyping of the concept has been experimentally validated in a mock up. → High temperature version of the ISI device has been made and tested in mock-up. Further experimentation is underway. → Simulation of different configurations of the device has been carried out with respect to reduced gap between main vessel and safety vessel for future FBRs. → Studies on wheel lining for the device have been carried out at 150 o C for better traction and payload capability. - Abstract: In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.

  3. The impact of the CO_2 separation system integration with a 900 MW_e power unit on its thermodynamic and economic indices

    International Nuclear Information System (INIS)

    Łukowicz, Henryk; Mroncz, Marcin

    2015-01-01

    This paper presents an analysis of the thermal cycle of a supercritical 900 MW_e condensing power plant which meets the “capture ready” requirements. The CO_2 separation method selected for the analysis is chemical absorption using MEA (monoethanolamine) or ammonia as sorbent. The indispensable scope of the turbine system upgrade necessitated by the incorporation of the carbon dioxide separation installation is proposed. The change in indices of the power unit operation after integration with the capture installation is presented for different variants of the retrofit. If MEA is used for carbon dioxide separation, the smallest drop in electric power can be observed in the case of hard coal for added stages at the intermediate pressure part outlet. In the case of lignite, the most favourable upgrade solution in terms of the smallest drop in electric power is elimination of one low pressure part and a backpressure turbine is added at the same time. If ammonia is used as sorbent, the best upgrade solution in terms of the smallest drop in electric power is the variant with more stages added at the IP part outlet, regardless of the type of fuel. An economic analysis is conducted for the proposed variants. - Highlights: • Impact of steam extraction on the turbine operation. • Adding more stages at the intermediate pressure part outlet. • Installing a backpressure turbine. • Eliminating one of the operating low pressure parts. • Economic assessment of proposed variants.

  4. Study of the muon spectrum at a depth 570 m.w.e. underground by means of the 100-ton scintillation detector

    International Nuclear Information System (INIS)

    Enikeev, R.I.; Zatsepin, G.T.; Korol'kova, E.V.; Kudryavtsev, V.A.; Mal'gin, A.S.; Ryazhskaya, O.G.; Khal'chukov, F.F.

    1988-01-01

    The experiment was carried out using the 100-ton apparatus at the Artemovsk Scientific Station of the Institute of Nuclear Research, USSR Academy of Sciences, located in a salt mine at a depth 570 m.w.e. underground. The spectrum of the energy release in the electromagnetic cascades which are generated by muons underground was measured. The electromagnetic and nuclear cascades were separated on the basis of the number of neutrons in these cascades. The spectrum of the energy release obtained is consistent with a spectrum of π and K mesons with γ/sub π//sub ,//sub K/ = 1.75 +- 0.08 for muon energies at sea level E 0 /sub μ/ >0.7 TeV. The experimental data recalculated to the vertical spectrum of muons at sea level agree with the results of other studies. Up to energies of about 100 TeV neither the spectrum of the primary cosmic rays nor the characteristics of the pA interaction undergo changes which could lead to an increase of γ/sub π//sub ,//sub K/ to a value exceeding 1.85

  5. Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF / B-VII.1

    Science.gov (United States)

    Saha, Uttiyoarnab; Devan, K.; Bachchan, Abhitab; Pandikumar, G.; Ganesan, S.

    2018-04-01

    The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA, ENDF / B-VII.1, wherein revisions were taken place in the new evaluations of basic nuclear data because of using the state-of-the-art neutron cross-section experiments, nuclear model-based predictions and modern data evaluation techniques. An indigenous computer code, computation of radiation damage (CRaD), is developed at our centre to compute primary-knock-on atom (PKA) spectra and displacement cross-sections of materials both in point-wise and any chosen group structure from the evaluated nuclear data libraries. The new radiation damage model, athermal recombination-corrected displacement per atom (arc-dpa), developed based on molecular dynamics simulations is also incorporated in our study. This work is the result of our earlier initiatives to overcome some of the limitations experienced while using codes like RECOIL, SPECTER and NJOY 2016, to estimate radiation damage. Agreement of CRaD results with other codes and ASTM standard for Fe dpa cross-section is found good. The present estimate of total dpa in D-9 steel of PFBR necessitates renormalisation of experimental correlations of dpa and radiation damage to ensure consistency of damage prediction with ENDF / B-VII.1 library.

  6. Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

    International Nuclear Information System (INIS)

    Su'ud, Zaki

    2008-01-01

    In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

  7. Hollow Fiber Membrane Contactors for CO2 Capture: Modeling and Up-Scaling to CO2 Capture for an 800 MWe Coal Power Station

    Directory of Open Access Journals (Sweden)

    Kimball Erin

    2014-11-01

    Full Text Available A techno-economic analysis was completed to compare the use of Hollow Fiber Membrane Modules (HFMM with the more conventional structured packing columns as the absorber in amine-based CO2 capture systems for power plants. In order to simulate the operation of industrial scale HFMM systems, a two-dimensional model was developed and validated based on results of a laboratory scale HFMM. After successful experiments and validation of the model, a pilot scale HFMM was constructed and simulated with the same model. The results of the simulations, from both sizes of HFMM, were used to assess the feasibility of further up-scaling to a HFMM system to capture the CO2 from an 800 MWe power plant. The system requirements – membrane fiber length, total contact surface area, and module volume – were determined from simulations and used for an economic comparison with structured packing columns. Results showed that a significant cost reduction of at least 50% is required to make HFMM competitive with structured packing columns. Several factors for the design of industrial scale HFMM require further investigation, such as the optimal aspect ratio (module length/diameter, membrane lifetime, and casing material and shape, in addition to the need to reduce the overall cost. However, HFMM were also shown to have the advantages of having a higher contact surface area per unit volume and modular scale-up, key factors for applications requiring limited footprints or flexibility in configuration.

  8. Solar thermal power stations for activities implemented jointly. The Theseus 50 MWe solar thermal power plant for the island of Crete, Greece

    Energy Technology Data Exchange (ETDEWEB)

    Brakmann, Georg [Fichtner, Stuttgart (Germany); Aringhoff, Rainer [Pilkington Solar International (United Kingdom); Cobi, Arend [PreussenElektra (Germany)

    1998-09-01

    THESEUS, the proposed commercial 50 MWe (net) Thermal Solar European Power Station for the Island of Crete is a solar hybrid plant with parabolic trough collectors and an advanced high efficiency Rankine reheat steam cycle. At the end of 1996 the DG XVII (Energy) of the European Commission has accepted the THERMIE application of the THESEUS consortium for the design phase. THESEUS reduces the required oil imports by 28 000 t/a, thereby saving the Greek economy every year 4 million ECU in foreign currency. During its 25 years technical lifetime 2.2 million tons of CO{sub 2} emissions will be avoided. Supply, construction, erection and operation of THESEUS creates 2 000 qualified employments (man-years). Because of the high manpower intensity of solar plants and their larger capital income from interest payments in contrast to the high fuel import intensity of fossil plants, THESEUS will generate larger tax revenues for Greece and for the supplier`s countries. The investment cost of THESEUS is some 135 million ECU. Even without any subsidies this would result in electricity generation cost of some 0.085 ECY/kWh, which is lower than the current average cost from the existing power plants of Crete. (author)

  9. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  10. Overall evaluation of combustion and NO(x) emissions for a down-fired 600 MW(e) supercritical boiler with multiple injection and multiple staging.

    Science.gov (United States)

    Kuang, Min; Li, Zhengqi; Liu, Chunlong; Zhu, Qunyi

    2013-05-07

    To achieve significant reductions in NOx emissions and to eliminate strongly asymmetric combustion found in down-fired boilers, a deep-air-staging combustion technology was trialed in a down-fired 600 MWe supercritical utility boiler. By performing industrial-sized measurements taken of gas temperatures and species concentrations in the near wing-wall region, carbon in fly ash and NOx emissions at various settings, effects of overfire air (OFA) and staged-air damper openings on combustion characteristics, and NOx emissions within the furnace were experimentally determined. With increasing the OFA damper opening, both fluctuations in NOx emissions and carbon in fly ash were initially slightly over OFA damper openings of 0-40% but then lengthened dramatically in openings of 40-70% (i.e., NOx emissions reduced sharply accompanied by an apparent increase in carbon in fly ash). Decreasing the staged-air declination angle clearly increased the combustible loss but slightly influenced NOx emissions. In comparison with OFA, the staged-air influence on combustion and NOx emissions was clearly weaker. Only at a high OFA damper opening of 50%, the staged-air effect was relatively clear, i.e., enlarging the staged-air damper opening decreased carbon in fly ash and slightly raised NOx emissions. By sharply opening the OFA damper to deepen the air-staging conditions, although NOx emissions could finally reduce to 503 mg/m(3) at 6% O2 (i.e., an ultralow NOx level for down-fired furnaces), carbon in fly ash jumped sharply to 15.10%. For economical and environment-friendly boiler operations, an optimal damper opening combination (i.e., 60%, 50%, and 50% for secondary air, staged-air, and OFA damper openings, respectively) was recommended for the furnace, at which carbon in fly ash and NOx emissions attained levels of about 10% and 850 mg/m(3) at 6% O2, respectively.

  11. Joint studies of LOF and TOP incidents for a 1300 MW(E) LMFBR using the computer codes SAS3D/EPIC and FRAX-2

    International Nuclear Information System (INIS)

    Leslie, R.; Billington, D.E.; Mann, J.E.

    1982-04-01

    The results of joint studies carried out for a 1300MW(E) LMFBR are described. The incidents examined were a slow TOP (3c/s) and a LOF (pump rundown with 9s flow halving time), both with failure to trip. For the TOP incident a benign outcome was predicted largely as a consequence of the prediction of clad failure near the top of the core. For the LOF incident highly energetic outcomes were not predicted for the reference case because the incident was terminated by disassembly (by fuel vapour pressure) in voided channels and failures in low-rated flooded channels with MFCI potential were not predicted. In the variant cases where MFCIs were predicted before shutdown, and rapid enough extension of the clad rips was allowed, low energetics were still predicted as a consequence of fuel sweepout. The strength of the MFCIs (as represented by a Cho-Wright treatment) does not appear to be an important factor but the results are dependent on the prediction of negative reactivity addition through fuel sweepout. The physical conditions obtaining at the time of fuel failure are such as to suggest that internal fuel motion following failure should not have an important effect on accident energetics, unless the development of the initial rip is delayed by several milliseconds. This is an area where only limited experimental evidence is available. Other areas of uncertainty are associated with the position of failure, of clad rip propagation and the influence of incoherency on the progression of the incident. Clad motion effects were shown not to influence accident energetics significantly for the reactor model considered. (author)

  12. Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

    Energy Technology Data Exchange (ETDEWEB)

    Foucher, L., E-mail: laurent.foucher@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Cousin, F.; Fleurot, J. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Brethes, S. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-CRI/SESUC, Cadarache, Saint-Paul-lez-Durance 13115 (France)

    2014-06-01

    In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool.

  13. Measurement of gas species, temperatures, coal burnout, and wall heat fluxes in a 200 MWe lignite-fired boiler with different overfire air damper openings

    Energy Technology Data Exchange (ETDEWEB)

    Jianping Jing; Zhengqi Li; Guangkui Liu; Zhichao Chen; Chunlong Liu [Harbin Institute of Technology, Harbin (China). School of Energy Science and Engineering

    2009-07-15

    Measurements were performed on a 200 MWe, wall-fired, lignite utility boiler. For different overfire air (OFA) damper openings, the gas temperature, gas species concentration, coal burnout, release rates of components (C, H, and N), furnace temperature, and heat flux and boiler efficiency were measured. Cold air experiments for a single burner were conducted in the laboratory. The double-swirl flow pulverized-coal burner has two ring recirculation zones starting in the secondary air region in the burner. As the secondary air flow increases, the axial velocity of air flow increases, the maxima of radial velocity, tangential velocity and turbulence intensity all increase, and the swirl intensity of air flow and the size of recirculation zones increase slightly. In the central region of the burner, as the OFA damper opening widens, the gas temperature and CO concentration increase, while the O{sub 2} concentration, NOx concentration, coal burnout, and release rates of components (C, H, and N) decrease, and coal particles ignite earlier. In the secondary air region of the burner, the O{sub 2} concentration, NOx concentration, coal burnout, and release rates of components (C, H, and N) decrease, and the gas temperature and CO concentration vary slightly. In the sidewall region, the gas temperature, O{sub 2} concentration, and NOx concentration decrease, while the CO concentration increases and the gas temperature varies slightly. The furnace temperature and heat flux in the main burning region decrease appreciably, but increase slightly in the burnout region. The NOx emission decreases from 1203.6 mg/m{sup 3} (6% O{sub 2}) for a damper opening of 0% to 511.7 mg/m{sup 3} (6% O{sub 2}) for a damper opening of 80% and the boiler efficiency decreases from 92.59 to 91.9%. 15 refs., 17 figs., 3 tabs.

  14. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  15. Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

    International Nuclear Information System (INIS)

    Foucher, L.; Cousin, F.; Fleurot, J.; Brethes, S.

    2014-01-01

    In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool

  16. Co-firing Bosnian coals with woody biomass: Experimental studies on a laboratory-scale furnace and 110 MWe power unit

    Directory of Open Access Journals (Sweden)

    Smajevic Izet

    2012-01-01

    Full Text Available This paper presents the findings of research into cofiring two Bosnian cola types, brown coal and lignite, with woody biomass, in this case spruce sawdust. The aim of the research was to find the optimal blend of coal and sawdust that may be substituted for 100% coal in large coal-fired power stations in Bosnia and Herzegovina. Two groups of experimental tests were performed in this study: laboratory testing of co-firing and trial runs on a large-scale plant based on the laboratory research results. A laboratory experiment was carried out in an electrically heated and entrained pulverized-fuel flow furnace. Coal-sawdust blends of 93:7% by weight and 80:20% by weight were tested. Co-firing trials were conducted over a range of the following process variables: process temperature, excess air ratio and air distribution. Neither of the two coal-sawdust blends used produced any significant ash-related problems provided the blend volume was 7% by weight sawdust and the process temperature did not exceed 1250ºC. It was observed that in addition to the nitrogen content in the co-fired blend, the volatile content and particle size distribution of the mixture also influenced the level of NOx emissions. The brown coal-sawdust blend generated a further reduction of SO2 due to the higher sulphur capture rate than for coal alone. Based on and following the laboratory research findings, a trial run was carried out in a large-scale utility - the Kakanj power station, Unit 5 (110 MWe, using two mixtures; one in which 5%/wt and one in which 7%/wt of brown coal was replaced with sawdust. Compared to a reference firing process with 100% coal, these co-firing trials produced a more intensive redistribution of the alkaline components in the slag in the melting chamber, with a consequential beneficial effect on the deposition of ash on the superheater surfaces of the boiler. The outcome of the tests confirms the feasibility of using 7%wt of sawdust in combination

  17. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  18. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  19. Influence of staged-air on airflow, combustion characteristics and NO(x) emissions of a down-fired pulverized-coal 300 MW(e) utility boiler with direct flow split burners.

    Science.gov (United States)

    Li, Zhengqi; Kuang, Min; Zhang, Jia; Han, Yunfeng; Zhu, Qunyi; Yang, Lianjie; Kong, Weiguang

    2010-02-01

    Cold airflow experiments were conducted to investigate the aerodynamic field in a small-scale furnace of a down-fired pulverized-coal 300 MW(e) utility boiler arranged with direct flow split burners enriched by cyclones. By increasing the staged-air ratio, a deflected flow field appeared in the lower furnace; larger staged-air ratios produced larger deflections. Industrial-sized experiments on a full-scale boiler were also performed at different staged-air damper openings with measurements taken of gas temperatures in the burner region and near the right-side wall, wall heat fluxes, and gas components (O(2), CO, and NO(x)) in the near-wall region. Combustion was unstable at staged-air damper openings below 30%. For openings of 30% and 40%, late ignition of the pulverized coal developed and large differences arose in gas temperatures and heat fluxes between the regions near the front and rear walls. In conjunction, carbon content in the fly ash was high and boiler efficiency was low with high NO(x) emission above 1200 mg/m(3) (at 6% O(2) dry). For fully open dampers, differences in gas temperatures and heat fluxes, carbon in fly ash and NO(x) emission decreased yielding an increase in boiler efficiency. The optimal setting is fully open staged-air dampers.

  20. The R and D D`s bearing test benches; Les bancs d`essais de paliers de la DER

    Energy Technology Data Exchange (ETDEWEB)

    Vialettes, J.M. [Service Ensembles de Production, Departement Machines, Direction des Etudes et Recherches, Electricite de France (EDF), 92 - Clamart (France)

    1997-01-01

    In power generation plants, rotating machines are involved in energy transformation processes and safety systems. The bearings supporting the rotors and the thrust bearings play a crucial role in the reliability of these machines. The phenomena encountered straddle several disciplines: hydrodynamics, tribology, thermomechanics, materials and vibrations in a specific environment, namely: thin fluid film, solid mechanical components and shaft rotation. Means of analysing the behaviour of these components (bearings and thrust bearings) have been developed and implemented. These consists of the EDYOS (Etude Dynamique des Organes de Supportage) code for dynamically studying bearing devices and several related bench tests. In reality, in order to understand the complex physical phenomena encountered in these components, it is vital to carry out analyses and experimental validations. Since these investigations cannot be carried out on actual machines, test benches have been built which can subject the sample bearings to the equivalent stresses. (author) 14 figs.

  1. Regulating low-NOx and high-burnout deep-air-staging combustion under real-furnace conditions in a 600 MWe down-fired supercritical boiler by strengthening the staged-air effect.

    Science.gov (United States)

    Kuang, Min; Wang, Zhihua; Zhu, Yanqun; Ling, Zhongqian; Li, Zhengqi

    2014-10-21

    A 600 MW(e) down-fired pulverized-coal supercritical boiler, which was equipped with a deep-air-staging combustion system for reducing the particularly high NOx emissions, suffered from the well-accepted contradiction between low NOx emissions and high carbon in fly ash, in addition to excessively high gas temperatures in the hopper that jeopardized the boiler's safe operations. Previous results uncovered that under low-NOx conditions, strengthening the staged-air effect by decreasing the staged-air angle and simultaneously increasing the staged-air damper opening alleviated the aforementioned problems to some extent. To establish low-NOx and high-burnout circumstances and control the aforementioned hopper temperatures, a further staged-air retrofit with horizontally redirecting staged air through an enlarged staged-air slot area was performed to greatly strengthen the staged-air effect. Full-load industrial-size measurements were performed to confirm the availability of this retrofit. The present data were compared with those published results before the retrofit. High NOx emissions, low carbon in fly ah, and high hopper temperatures (i.e., levels of 1036 mg/m(3) at 6% O2, 3.72%, and about 1300 °C, respectively) appeared under the original conditions with the staged-air angle of 45° and without overfire air (OFA) application. Applying OFA and reducing the angle to 20° achieved an apparent NOx reduction and a moderate hopper temperature decrease while a sharp increase in carbon in fly ash (i.e., levels of 878 mg/m(3) at 6% O2, about 1200 °C, and 9.81%, respectively). Fortunately, the present staged-air retrofit was confirmed to be applicable in regulating low-NOx, high-burnout, and low hopper temperature circumstances (i.e., levels of 867 mg/m(3) at 6% O2, 5.40%, and about 1100 °C, respectively).

  2. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts; Pompes primaires 93 D des tranches de 900 MW. Conditions thermo-hydrauliques d`amorcage des fissures d`arbres

    Energy Technology Data Exchange (ETDEWEB)

    Bore, C.

    1995-12-31

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a `viscosity pump` phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. (Abstract Truncated)

  3. French experience on renewing I and C systems in NPPs. Feedback from assessing nuclear instrumentation system (RPN) refurbishment at French CP0-series plants

    International Nuclear Information System (INIS)

    Elsensohn, O.; Fradet, F.; Peron, J.C.; Soubies, B.

    2003-01-01

    In 1996, the utility operating France's nuclear power plants launched feasibility studies for the refurbishment of the nuclear instrumentation system (RPN classed category A) installed in its CPO-series (900 MWe) units. The system was ultimately upgraded with digital I and C system, using a SPINLINE 3 platform. This article describes feedback from an evaluation conducted on the refurbishment by the Institute of Radiological Protection and Nuclear Safety (IRSN), technical support arm of the Directorate General for Nuclear Safety and Radiological Protection (DGSNR). The study begins with a historical overview of the refurbishing operation, then discusses the IRSN assessment method and the lessons learned from this first major revamp of an I and C system in the French nuclear reactor series. Based on its previous experience in evaluating I and C systems for P4/P'4 (1300 MWe) and N4 (1450 MWe) plants and to account for the first-ever aspect of such an upgrade, IRSN partitioned its assessment into four phases. This approach enabled taking into account the impact of RPN refurbishment at every level - system, hardware and qualification, software, operation, onsite requalification, health physics, fire protection and human factors. All six units in the CPO series have now been equipped with the new digital RPN. (authors)

  4. Supply of clean water to the bearings and mechanical seals of the backup pumps; Alimentation en eau propre des paliers et garnitures mecaniques des pompes de sauvetage

    Energy Technology Data Exchange (ETDEWEB)

    Jolas, C. [Department Machines, Service Ensembles de Production, Direction des Etudes et Recherches, Electricite de France (EDF), 92 - Clamart (France)

    1997-01-01

    The purpose of the backup pumps is to cool the primary circuit and pressurised water reactor containment in the case of a primary cooler loss accident. The water taken in by these pumps in the case of accident is loaded with solid particles. In order to ensure correct operation of the bearings and mechanical seals of these machines, they must be supplied with clean water. In other words, the solid particles must be removed from the water intake. Manufacturers generally use cyclonic separators to achieve this. (author) 5 refs., 14 figs.

  5. Using the drinking water of the Municipality of St-Gingolph, Switzerland for power generation - Preliminary study; Turbinage des eaux potables de la commune de St-Gingolph. 1{sup ere} partie: Source de la Tine - Palier superieur. 2{sup e} partie: Sources de la Tine et de Clarive - Palier inferieur

    Energy Technology Data Exchange (ETDEWEB)

    Moukhliss, H.

    2008-07-15

    This illustrated report subdivided in two parts presents the preliminary study of a project aiming at installing two small-scale hydroelectric schemes on the drinking water supply network of the Municipality of St-Gingolph, Switzerland. The upper-spring (La Tine) water can be used in a 44 kW hydro power scheme (elevation difference: 121 m; water volume through the turbine: 1,240,000 m{sup 3}/y). This water could be added to the one from the Clarive spring, enabling the installation a second hydro power scheme with a power of 291 kW (elevation difference: 337 m; water volume through the turbine: 2,740'000 m{sup 3}/y). The estimated cost of the produced electricity is attractive: CHF 0.16/kWh for the La Tine scheme; CHF 0.11 for the Clarive scheme. The author recommends the construction of both schemes.

  6. The nuclear instrumentation system of the French 1400 MWe reactors

    International Nuclear Information System (INIS)

    Bourgerette, A.; Mauduit, J.P.

    1993-01-01

    The nuclear instrumentation systems in power reactors in France have made considerable advances thanks to technological progress. The appearance of an integrated digital protection system (SPIN) and the extension of digital techniques have considerably improved performance and operating flexibility. Working on the basis of technology developed jointly with the Nuclear Electronics and Instrumentation Department at the French Atomic Energy Commission (CEA), Framatome and Merlin Gerin have designed the new nuclear instrumentation system for 1400 MW reactors. (authors). 4 figs

  7. EPR by Areva. EPR the 1600+ MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system.

  8. Acceptance test for 900 MWe PWR unit replacement steam generators

    International Nuclear Information System (INIS)

    Gourguechon, B.

    1993-01-01

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG's differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs

  9. Predisposal of Radioactive Waste from NPP 1000 MWe

    International Nuclear Information System (INIS)

    Suryantoro

    2007-01-01

    Predisposal of radioactive waste from NPP 1000 MW which was planned to be operated in 2016 has been conducted. In this study NPP applying PWR type was assumed. This assessment comprises all aspects of radioactive waste coming from NPP. One through cycle was chosen consequently no reprocessing step will be conducted. The assessment shows that technologically all radioactive waste treatment process rising from NPP operation has similarities to the existing radioactive waste process conducted by RWI which has lower scale of waste amount. (author)

  10. Seismic Margin of 500MWe PFBR Beyond Safe Shutdown Earthquake

    International Nuclear Information System (INIS)

    Sajish, S.D.; Chellapandi, P.; Chetal, S.C.

    2012-01-01

    Summary: • Seismic design aspects of safety related systems and components of PFBR is discussed with a focus on reactor assembly components. • PFBR is situated in a low seismic area with a peak ground acceleration value of 0.156 g. • The design basis ground motion parameters for the seismic design are evaluated by deterministic method and confirmed by probabilistic seismic hazard analysis. • Review of the seismic design of various safety related systems and components indicate that margin is available to meet any demand due to an earthquake beyond SSE. • Reactor assembly vessels are the most critical components w.r.t seismic loading. • Minimum safety margin is 1.41 for plastic deformation and 1.46 against buckling. • From the preliminary investigation we come to the conclusion that PFBR can withstand an earthquake up to 0.22 g without violating any safety limits. • Additional margin can be estimated by detailed fragility analysis and seismic margin assessment methods

  11. Extended Station Blackout Analysis for VVER-1000 MWe Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, A. J.; Rao, R. S.; Lakshmanan, S. P.; Gupta, A., E-mail: avinashg@aerb.gov.in [Atomic Energy Regulatory Board, Mumbai (India)

    2014-10-15

    Post Fukushima, the plant behaviour for an extended station black-out (ESBO) scenario with only passive system availability for about 7 days has become imperative. Thermal hydraulic analysis of ESBO with the availability of passive heat removal system (PHRS), passive first stage and second stage hydro accumulators were carried out to demonstrate the design capabilities. Two different cases having primary leak rates of 2.2 tons/hr and 6.6 tons/hr were analyzed to study sustenance of natural circulation. For the study, out of 4 PHRS trains, one PHRS train was assumed to be in failure mode. The objective here is to predict the core cooling capability for a period of 7 days under ESBO conditions with the available water inventories from first and second stage hydroaccumulators only. Over simplified energy balance studies cannot ascertain sustenance of natural circulation in the primary system, steam generators (SGs) and PHRS. The analysis was carried out by using system thermal hydraulic safety code RELAP5/SCDAP/MOD 3.4. It is inferred that the inventory in the first stage accumulators and second stage accumulators compensate the leak and decay heat is removed effectively with the help of passive heat removal systems. It is also observed that even after 7 days of ESBO a large inventory is still available in the second stage accumulators and the primary system remains subcooled. (author)

  12. Hybrid simulation of a 900 Mwe nuclear plant

    International Nuclear Information System (INIS)

    Constantieux, Thierry; Deat, Max.

    1979-01-01

    To analyse the effects on PWRs of transients originating from the network, specific means of calculation must be elaborated. One of them which was conceived and set up on a hybrid computer by FRAMATOME and the FRENCH ATOMIC ENERGY COMMISSION, is described in this paper. The method chosen to validate this code is fairly original, since it consisted in carrying out a long duration test on a plan and in simulating this on the hybrid computer; then in carefully comparing the recorded data of the test with the results of the simulation. The quality of the results, thus obtained shows that a relatively unsophisticated model is able to give a good idea of actual process behavior, but only if the types of transients to be studied with the code are well identified before its elaboration

  13. Multivariable controller for a 600 MWe CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Mensah, S.

    1982-11-01

    The problems of designing a multivariable regulator for a nuclear power station of the Gentilly-2 type are studied. A reduced model, G2LDM, linearized around steady state operating conditions, is derived from the non-linear model G2SIM. The resulting linear model is described by state-space equations. Good agreement is demonstrated between the transient responses of both models. Properties of G2LDM are assessed by performing controllability and observability tests, cyclicity and rank tests, and eigenanalysis. A comprehensive set of application-orinented algorithms which allow multivariable controller design with closed-loop pole-assignment techniques are implemented in a computer-aided design package via several modules. A general scheme for the implementation of a multivariable controller in G2SIM is designed, and simulation tests show satisfactory performance of the controller [fr

  14. EPR by Areva. EPR the 1600+ MWe reactor

    International Nuclear Information System (INIS)

    2008-01-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

  15. Design of the control room of the N4-type PWR: main features and feedback operating experience; La salle de commande du palier N4: principales caracteristiques et retour d'experience d'exploitation

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouton, J.M.; Guillas, J.; Nougaret, Ch. [Electricite de France (EDF/DPN/CAPE), 93 - Saint-Denis (France)

    2004-07-01

    This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

  16. Modular plants with high power gas engines (1 to 30 MWe); Centrales modulaires a moteurs gaz de forte puissance (de 1 a 30 MWe)

    Energy Technology Data Exchange (ETDEWEB)

    Haushalter, J. [Wartsila (France)

    1997-12-31

    This paper is a series of transparencies about the high power gas engines manufactured by Waertsilae NSD Corporation company. The first par recalls the NO{sub x} and CO air pollution regulations worldwide, the German TA-Luft standards and the French 2910 by-law according to the engine type (2 and 4 stroke, dual-fuel, natural gas, LPG, others..) and to the type of pollutants (NO{sub x}, dusts, SO{sub 2}, CO, noise..). The second part presents the Waertsilae NSD Corporation concept of gas-fueled spark ignition engines (Otto cycle, emissions, performances, technology, fuel system, combustion optimization, fuel-air ratio regulation, pollution control equipment) and of the `pure energy` global concept of plants. (J.S.)

  17. Modular plants with high power gas engines (1 to 30 MWe); Centrales modulaires a moteurs gaz de forte puissance (de 1 a 30 MWe)

    Energy Technology Data Exchange (ETDEWEB)

    Haushalter, J. [Wartsila NSD (France)

    1997-12-31

    After a review of pollution regulations in France and Europe for high capacity combustion plants, the Wartsila NSD spark ignition combustion system, using natural gas, is presented: the air-gas mixture in the combustion chamber is very weak (lambda is around 2-2.2) and its ignition is completed by the flame exiting the pre-chamber containing a stoichiometric mixture, and the spark plug. The temperature is decreased thus lowering the NOx emission level. The combustion system is integrated in the Pure Energy global concept (cogeneration plants, etc.) from Wartsila

  18. Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

    International Nuclear Information System (INIS)

    Nawathe, Shirish; Umashankari, P.; Balakrishnan, Kamala; Mahajan, S.C.; Kakodkar, A.

    1991-01-01

    A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

  19. Leibstadt: a 950-MW(e) BWR/6 Mark-III in commercial operation

    International Nuclear Information System (INIS)

    Fischer, P.U.

    1985-01-01

    It may be somewhat premature to report on a plant that started up in 1984 as the first of General Electric's (GE's) BWR/6 Mark-III plants in the Western Hemisphere and commenced commercial operation on December 15, 1984. The theme of the session certainly applies to the overall Swiss nuclear program and the search for excellence has been our ambition out of economic and energy supply necessities. Leibstadt came on line just in time to cover the needs of the Swiss consumers during the winter of 84/85. It has provided reliable service from the outset and operated during the extreme European cold wave in January 1985 without interruption. In 1985 the plant is expected to cover approx.15% of the electricity needs of Switzerland. The encouraging start of commercial operation gives hope that with time Leibstadt will be able to approach the capacity factors of the other four Swiss nuclear power stations, which in 1984 were between 88.4 and 90.3%

  20. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

    2005-08-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

  1. Atucha II: building a 745 MWe pressure-vessel PHWR in the Argentine

    International Nuclear Information System (INIS)

    Leibovich, H.; Coll, J.; Backhaus, K.

    1982-01-01

    The history of the Comision Nacional de Energia Atomica and its involvement with the development in Argentina of a nuclear power plant manufacturing technology is described. The way in which technology transfer from German industry has enabled increasing Argentine participation is outlined. The formation of the joint Argentine/German engineering company ENACE in 1981 will promote this process. (U.K.)

  2. Liquid poison injection system (LPIS) for KAIGA 1, 2 and RAPP 3, 4, 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Soni, K.L.; Aparna; Mohan, L.R.; Nema, M.K.; Mahajan, S.C.

    1997-01-01

    LPIS is the modified version of existing Bulk Addition Mode (BAM) of the Automatic Liquid Poison Addition System (ALPAS). This BAM mode of ALPAS serves as slow acting supplementary shutdown system to PSS or SS as the case may be and hence it becomes a part of reactor safety system. The system (LPIS) has been envisaged to perform similar functions of ALPAS mode BAM. Now this is a self standing system. This feature of the system eliminates need of Gravity Addition of Boron System (GRAB). As the concept evolved is to be introduced for first time for the power reactors (from KAIGA-1, 2 and RAPP-3, 4 onwards), it has become necessary to check and verify its working by carrying out the necessary experimental and developmental work

  3. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  4. Evaluation of 450-MWe BGL GCC power plants fueled with Pittsburgh No. 8 coal

    International Nuclear Information System (INIS)

    Pechtl, P.A.; Chen, T.P.; Thompson, B.H.; Greil, C.F.; Niermann, S.E.; Jandrisevits, M.

    1992-11-01

    In this study, a conceptual design and cost estimate were developed for a nominal 450 MW integrated gasification combined cycle plant using the British Gas/Lurgi slagging gasification process. The present study is a design update of a previous study (EPRI Report AP-6011). The major design improvements incorporated include use of the latest GE 7F gas turbine rating, integrating the air separation plant with gas turbine, use of fuel gas saturation for NO x control, use of treated gasifier waste water as makeup water for the fuel gas saturation, and several process changes in the acid gas removal and sulfur recovery areas. Alternate design options for feeding the excess coal fines to the gasifier, treating the gasifier waste water, and the use of conventional air separation without integration with gas turbine were evaluated. The design improvements incorporated were found to increase significantly the overall plant efficiency and reduce the cost reported in the previous study. The various design options evaluated were found to have significant impacts on the plant efficiency but negligible impacts on the cost of electricity

  5. 400-MWe Consolidated Nuclear Steam System (CNSS). 1200-MWt Phase 2A interim studies

    International Nuclear Information System (INIS)

    1978-09-01

    The Phase 2A interim studies of the Consolidated Nuclear Steam System (CNSS) consisted of a number of separate task studies addressing the design concepts developed during the Phase 1 study reported in BAW--1445. The purpose of the interim studies was to better establish overall concept feasibility from both a hardware and economic standpoint, to make modification and additions to the design where appropriate, and to understand and reduce the technical risks in critical areas of the design. The work on these task studies included input from Barberton, Mt. Vernon, and the Alliance Research Center as well as United Engineers and Constructors (UE and C). The UE and C work was carried out under a separate DOE contract

  6. A final report on stress analysis of seal disc of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Chawla, D.S.; Dutta, B.K.; Kushwaha, H.S.; Kakodkar, A.; Sanatkumar, A.

    1989-01-01

    In Pressurised Heavy Water Reactor (PHWR) on-power refuelling is done by use of fuelling machine. Before refuelling, sealing plug assembly is removed from the end-fitting of the coolant channel and after refuelling the sealing plug is reinstalled back in to the end-fitting. The seal disc is a part of sealing plug assembly. Its function is to create sealing action for the heavy water inside the coolant channel. A systematic developmental work is done to arrive at a final configuration of the seal disc. This is done to minimise the stresses in the body of the seal disc and at the same time to obtain required seating reaction to avoid heavy water leakage. It is observed that stresses computed for the final configuration by linear elastic analysis are more than the allowable value as per ASME Section III, Division 1. This calls for elasto-plastic analysis to find out collapse load to satisfy ASME codal limits as per special provision of NB-3228.1 (1986). The elasto-plastic analysis showed that the seal disc meets ASME codal limits for all stages of loading. (author). 8 refs., 2 tabs., 7 figs

  7. Development of channel inspection and gauging apparatus for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Parulkar, S.K.; Taneja, R.; Taliyan, S.S.; Singh, Manjit; Govindarajan, G.

    1992-01-01

    Channel inspection and gauging apparatus is being developed to enable in-service channel inspection and gauging. Phase I apparatus to measure annular gap between pressure tube and calandria tube in a dry channel has been developed. The apparatus consists of a gauging head and a drive mechanism. The gauging head utilities an eddy current probe to measure the annular gap between pressure tube and calandria tube and an ultrasonic sensor to measure the wall thickness of the pressure tube. The output signal of the eddy current probe needs to be corrected for the effect of pressure tube wall thickness variation. This paper gives the details of the above apparatus. The results of calibration tests at mock-up station are presented. The paper outlines the program for the phase-wise development of Channel Inspection and Gauging Apparatus for use in heavy water filled channels without their isolation from PHT and draining. The final apparatus will have the facilities for ultrasonic flaw detection, ultrasonic gauging to measure pressure tube diameter and wall thickness, an inclinometer to measure slope and sag of pressure tube and eddy current probe for the measurement of annular gap between pressure tube and calandria tube. (author). 6 figs

  8. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh

    2005-01-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  9. 10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-25

    Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

  10. The techno-economic optimization of a 100MWe CSP-desalination plant in Arandis, Namibia

    Science.gov (United States)

    Dall, Ernest P.; Hoffmann, Jaap E.

    2017-06-01

    Energy is a key factor responsible for a country's economic growth and prosperity. It is closely related to the main global challenges namely: poverty mitigation, global environmental change and food and water security [1.]. Concentrating solar power (CSP) is steadily gaining more market acceptance as the cost of electricity from CSP power plants progressively declines. The cogeneration of electricity and water is an attractive prospect for future CSP developments as the simultaneous production of power and potable water can have positive economic implications towards increasing the feasibility of CSP plant developments [2.]. The highest concentrations of direct normal irradiation are located relatively close to Western coastal and Middle-Eastern North-African regions. It is for this reason worthwhile investigating the possibility of CSP-desalination (CSP+D) plants as a future sustainable method for providing both electricity and water with significantly reduced carbon emissions and potential cost reductions. This study investigates the techno-economic feasibility of integrating a low-temperature thermal desalination plant to serve as the condenser as opposed to a conventional dry-cooled CSP plant in Arandis, Namibia. It outlines the possible benefits of the integration CSP+D in terms of overall cost of water and electricity. The high capital costs of thermal desalination heat exchangers as well as the pumping of seawater far inland is the most significant barrier in making this approach competitive against more conventional desalination methods such as reverse osmosis. The compromise between the lowest levelized cost of electricity and water depends on the sizing and the top brine temperature of the desalination plant.

  11. Sizing of ion exchange column for PHT purification system of 500 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Goswami, S; Sharma, A K; Bapat, C N; Sharma, V K [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Strict chemistry control is required on the heavy water coolant in the primary loop of the PHWR to ensure stability of oxide corrosion film and to prevent corrosion build up on the surfaces. This is achieved by maintaining PD value of PHT coolant at about 10.45 and sp. conductivity between 20 to 40 mhos/cm by ion exchange column and filtration of coolant. This paper deals with the sizing and selection of bed height of ninth column from different methods. (author). 7 refs., 6 figs.

  12. Systems Analysis of a Fast Steam-Cooled Reactor of 1000 MW(E)

    Energy Technology Data Exchange (ETDEWEB)

    Smidt, D.; Frisch, W.; Hofmann, F.; Moers, H.; Schramm, K.; Spilker, H. [Institut fuer Reaktorentwicklung, Kernforschungszentrum, Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Kiefhaber, E. [Institut fuer Neutronenphysik und Reaktortechnik Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1968-05-15

    The Karlsruhe design of a steam-cooled fast reactor (Dl) has been the subject of a systems analysis. Here the dependence of fuel inventory, breeding ratio, rating, core geometry and plant efficiency on coolant pressure, and coolant temperature has been studied for two different rod powers. The effect of artificial surface roughness has been investigated. For some configurations the resulting fuel-cycle and capital costs have been determined and discussed. The main influence results from pressure. The lower pressure allows for higher breeding ratios, but lower efficiencies and vice versa. From this the fuel-cycle costs show an optimum at around 150 atm abs. The capital costs on the other side decrease with pressure. The over-all optimum of the power generating costs for the presently studied parameter range is at about 170 atm abs., a coolant outlet temperature of 540 Degree-Sign C and a rod power of 420 W/cm. Artificial roughness (boundary layer type) leads for a required system pressure and outlet temperature to a larger coolant volume fraction and, therefore, to reduced breeding ratios but higher efficiencies. As another part of the work some stability characteristics of the cores were studied. The dependence of the core stability on the varied parameters is shown. (author)

  13. 400-MWe consolidated nuclear steam system (CNSS): 1200-MWt/conceptual design

    International Nuclear Information System (INIS)

    1977-06-01

    A 1200-MWt consolidated nuclear steam system (CNSS) conceptual design is described. The concept, derived from nuclear merchant ship propulsion steam systems but distinctly different from those systems in detail, incorporates the steam generators within the reactor pressure vessel. This configuration eliminates primary coolant circulating piping external to the reactor pressure vessel since the primary coolant circulating pumps are mounted in the pressure vessel head. So arranged, the maximum piping break that must be assumed is that of the pressurizer surge line, which is substantially smaller than a primary coolant circulating line. A fracture of the pressurizer surge line would result in substantially lower mass and energy release rates of the primary coolant during the assumed loss-of-coolant accident. This in turn makes practical a pressure-suppression containment rather than the ''dry'' containment commonly used for pressurized water reactors

  14. Expert system for the investigation of safety system availability on a 900 MWe PWR

    International Nuclear Information System (INIS)

    Chauliac, C.; Deplanque, B.; To, L.H.

    1988-01-01

    A computer program of the expert system type would appear to be an elegant and effective tool for rapid diagnosis of safety system availability in accident situations. The expert system developed for this purpose by the Institut de Protection et de Surete Nucleaire (Institute for Nuclear Safety and Protection) has been described in this paper; its logic process has been examined in detail and illustrated by means of two examples. In its present form, this expert system monitors the availability of 21 main systems. In its final form (1989), 37 main systems will be tested. It will then include descriptions of between 1500 and 2000 objects and will utilize about 1000 rules. It will be run (as is presently the case) in a workstation with windowing facilities and graphic result displays which provide the highest degree of user-friendliness

  15. Design of multivariable controller for a 600 MWe CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Mensah, S.; McMorran, P.D.

    1982-04-01

    This paper reports the results of a case study on the design of a multivariable regulator for a nuclear power station of the Gentilly-2 type. In this study, a design model was derived by simplifying and linearizing equations in the G2SIM non-linear model. Open-loop simulation showed good agreement between transient responses of both models. After a critical review of multivariable design techniques, the authors explored pole shifting with output feedback. A comprehensive set of application-oriented algorithms for closed-loop pole shifting, implemented via modules in the MVPACK computer-aided design package were derived. A controller was designed for the linear model, then implemented on the non-linear simulation. After adjustment of controller gains, mainly in the dynanamic section of the feedback, simulation results showed that the performance of the multivariable controller on G2SIM is satisfactory. The results demonstrate the relative superiority of the multi-variable controller over the existing conventional controller

  16. General Atomic's 1500-MWe high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Waage, J.M.; Goodjohn, A.J.; Crimmins, D.H.; Wessman, G.L.

    1974-01-01

    Information is presented concerning the specifications for the plant layout, nuclear steam supply system, core and major components, prestressed concrete pressure vessel, control system, and safety systems. (U.S.)

  17. Radiological Consequences Analysis for Abnormal Condition on NPPs 1000 MWe by Using Radcon Model

    International Nuclear Information System (INIS)

    Pande Mande Udiyani; Sri Kuntjoro

    2009-01-01

    The operation of NPPs (Nuclear Power Plants) in Indonesia to anticipates rare of energy will generate various challenges, especially about NPPs safety. So installation organizer of nuclear must provide scientific argument to safety NPPs, one of them is by providing document of safety analysis. Calculation of radiological consequences after abnormal condition applies on generic PWR-1000 power reactor. Calculation is done by using program package RadCon (Radiological Consequences Model), with postulate condition is based on DBA (Design Basis Accident). Calculation of dispersion of radionuclide concentration is using PC-COSYMA as input data for RadCon. Simulation for radiological consequences analysis uses by site data sample. Analysis result shows that maximum receiving of internal - externals radiological consequence for short term and long-term below 1 km radius area is below the limit acceptably effective dose for a member of the public as a result of an accident which should not exceed 5 mSv (ICRP 1990). (author)

  18. MHD repowering of a 250 MWe unit of the TVA Allen Steam Plant

    International Nuclear Information System (INIS)

    Chapman, J.N.; Attig, R.C.

    1992-01-01

    In this paper coal fired MHD repowering is considered for the TVA Allen Steam Plant. The performance of the repowered plant is presented. Cost comparisons are made of the cost of repowering with MHD versus the cost of meeting similar standards by installing scrubbers and selective catalytic NO x reduction (SCNR). For repowering of a single 250 MW e unit, the costs favor scrubbing and SCNR. If one considers a single repowering of all three 250 MW e units by a single MHD topping cycle and boiler, MHD repowering is more economical. Environmental emissions from the repowered plant are estimated

  19. Co-firing Coal and Straw in a 150 MWe Utility Boiler: Deposition Propensities

    DEFF Research Database (Denmark)

    Andersen, Karin Hedebo; Hansen, Peter Farkas Binderup; Wieck-Hansen, Kate

    1996-01-01

    In order to meet a 20 % reduction in CO2 emissions, based on 1988 levels, by the year 2005, the Danish Government has committed the power companies in Denmark to burn 1.2 million tons of straw per year from the year 2000. A conventional pf-fired boiler at the Danish Power Company Midtkraft has been...... on temperature controlled probes and in-situ measurements of flue gas temperature, flue gas compositions including alkali metal concentrations and particle and aerosol loading and composition. The study is carried out as collaborative research projects between the Midtkraft Power Company and the Combustion...

  20. The main security problems encountered in the definition of the PWR 1300 MWe stage

    International Nuclear Information System (INIS)

    Guimbail, H.; Auvergnon, F.

    1980-01-01

    The main problems of security encountered in the definition of the PWR 1300 stage originated from the necessity to ensure the continuity of a stage with the preceding one, the overture to technical progress, the acceptance of responsibility for the gaines experience and the control of cost prices. A few examples show how desirable it is for a project to be managed with the design and construction rules stabilized as early as possible [fr

  1. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

  2. Flue gas emissions from gas-fired cogeneration units <25 MWe

    International Nuclear Information System (INIS)

    Nielsen, M.; Wit, J. de

    1997-01-01

    A total of 900 MW e gas driven combined heat and power (CHP) has now been established in Denmark based on gas engines and gas turbine units less than 25 MW e each. Of the 900 MW e approx. 750 MW e are based on gas engines. Biogas is used as fuel for some 32 MW e of these. Emission limits for NO x and CO are 650 mg/nm 3 (ref. 5% O 2 and electrical efficiency 30% LCV). There is at present no limit for unburned hydrocarbons (UHC) for gas engines or gas turbines. The average emission of unburned hydrocarbons for the Danish gas engine driven CHP units is equal to approx. 3,5% of the fuel used. It is the target of this report to provide the basis for evaluating the planned UHC limit and possible adjustments of the present limit for NO x emission. The average NO x emission from gas turbines slightly exceeds the NO x emission from gas engines. This is due to a number of older gas turbines. Modern gas turbines can achieve significantly lower NO x emission compared to engines. The NO x emission from biogas driven engines is significantly higher than that of natural gas driven units. This is mainly due to NO x -unfavourable engine settings and the use of older units, as there are no legislation concerning NO x emission for the majority of these biogas driven units. The emission of CO and UHC is lower from gas turbines than from gas engines. The NO x emission can be reduced by SCR Catalyst systems. In Denmark 3 gas engine installations use this commercially available technology. Oxidation catalyst for UHC reduction at modern gas engine installations has proven relatively unsuccesful in Denmark until now. Only limited reductions are achieved and many catalysts are toxificated in less than 100 hours of operation. However, long-term field testing of promising UHC reducing catalysts is now being made. UHC reduction by incineration is at the prototype stage. No such plant has yet been set up in Denmark. (Abstract Truncated)

  3. Design of condenser for 500 MWe pressurised heavy water reactors (PHWRs) - a case study

    International Nuclear Information System (INIS)

    Agarwal, N.K.; Subbarao, A.; Chaudhary, K.

    1996-01-01

    Condenser forms the major heat sink in the power plants. In recent years, power plant availability and performance have become great concern to the industry. The detailed design of the condenser and its associated cooling water (CW) system require careful optimisation of parameters which include material selection, cooling water flow rate, condenser surface areas, turbine exhaust pressures etc. This is required to produce a design offering maximum efficiency and reliability and minimum maintenance. The various parameters involved in condenser design are discussed. 5 refs., 1 fig

  4. Lessons learned on shrunk-on disks turbines of the French 900 MWe PWR

    International Nuclear Information System (INIS)

    Calle, P.; Dordonat, M.; Dury, J.P.; Pages, G.; Gras, J.M.; Vaillant, F.; Fillon, R.; Aubry, P.; Bocquet, F.

    1990-01-01

    The LP shrunk-on discs of the 900 MW turbines have been submitted to 2 types of damage: - the fretting-fatigue which affects the housings of anti-rotation keys, of end discs in particular (discs 7 and 14); - the stress-corrosion of the so-called hot discs (3, 4, 5 and the symmetrical ones). Since the current situation of the park is not critical, a policy of revamping could be implemented in order to give the equipment an increased service life without changing the technology. In particular, two kinds of provisions have been taken to fight against the stress-corrosion: on the one hand, protection of the sensitive areas by performing surface treatments and on the other hand, application of constructive modifications intended to avoid containment areas. Besides, the studies associated with these solutions (nickel - plating, shot-blasting) have revealed that they have no negative effects on the corrosion resistance of the base steel. This policy which results from a close collaboration between GEC ALSTHOM and EDF has enabled to avoid any unavailability [fr

  5. A 100 MWe Advanced Sodium-cooled Fast Reactor (AFR-100)

    International Nuclear Information System (INIS)

    Grandy, C.; Kim, T.K.; Jin, E.

    2013-01-01

    • AFR-100 Design development is continuing in the U.S.; • Various innovations are included in the design to understand their feasibility; • Engineering and safety analyses have been performed that demonstrate the inherent safety characteristics of the AFR-100 design during severe accidents; • R&D is being performed on a number of the innovations such as advanced materials, compact fuel handing system, advanced energy conversion system, advanced core design, etc

  6. Safety analysis of the critical facility for AHWR and 500 MWe PHWR

    International Nuclear Information System (INIS)

    Pushpam, Neelima Prasad; Arvind Kumar; Srivenkatesan, R.

    2002-01-01

    Full text: The initiating event for the design basis reactivity accident is the uncontrolled moderator pump up at criticality. This uncontrolled pump up transient is considered to be the enveloping scenario and has been analysed for the reference core with AHWR fuel using point kinetics model. It is the most reactive core among the three with a small b value (core average b= 5.96 mk). The maximum pump rate in critical facility is limited to 300 litres/min which corresponds to the maximum rate of reactivity addition about 0.1 mk/sec. On any reactor trip 6 shut-off rods are inserted into the core along with partial moderator dumps. The reactor is provided with independent safety and regulating channels (SC and RC) to monitor reactor neutronic power and initiate trip at different power levels. After the reactor trips five of the six fast acting shut-off rods (maximum worth rod is unavailable) fall under gravity and at the same time moderator dump is initiated. We have considered shut off rods and moderator dump as two independent shutdown systems. The analysis shows that even if the reactor trips at the high power at 550 watt ignoring the earlier trips, the fuel temperature does not rise beyond 50 degC and the total energy released is less than 20 kW. We also analysed the transients due to uncontrolled withdrawal of absorber rod. In this case also we found that the fuel temperature became ∼54 degC and the total energy release was about 25 kW. The fuel can withstand this temperature. This shows that reactor is safe

  7. Zero-discharge wastewater treatment facility for a 900-MWe GCC power plant

    International Nuclear Information System (INIS)

    Rosain, R.M.; Dalan, J.A.

    1992-05-01

    Florida Power and Light desires to examine the prospect of achieving zero liquid discharge from the gasification area of their proposed 900-MW coal gasification-combined cycle (GCC) power plant expansion at the Martin station. This report provides information about the technologies available, cost, and process selection methods, and recommends a preferred system for achieving zero liquid discharge from the gasification block. The recommended system consists of primary clarification and vapor compression evaporation, followed by carbon adsorption post-treatment of the evaporator distillate. Dry solids are produced from the evaporator concentrate with a crystallizer/centrifuge combination. The system recovers 99 percent of the wastewater as pure distillate vater. The predicted capital cost for the 265-gpm system is $12.5 million; the predicted operating costs are $18.60/1000 gallons. Both costs are in 1990 dollars. Promising treatment technologies to examine for future designs are cooling tower treatment and freeze crystallization

  8. 400-MWe Consolidated Nuclear Steam System (CNSS). 1200-MWt Phase 2A interim studies. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The Phase 2A interim studies of the Consolidated Nuclear Steam System (CNSS) consisted of a number of separate task studies addressing the design concepts developed during the Phase 1 study reported in BAW--1445. The purpose of the interim studies was to better establish overall concept feasibility from both a hardware and economic standpoint, to make modification and additions to the design where appropriate, and to understand and reduce the technical risks in critical areas of the design. The work on these task studies included input from Barberton, Mt. Vernon, and the Alliance Research Center as well as United Engineers and Constructors (UE and C). The UE and C work was carried out under a separate DOE contract.

  9. Seismic analysis for safety related structures of 900MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  10. A comparative neutronic analysis of 150MWe TRU burner according to the coolant alteration

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic analysis has been conducted for the small TRU burner according to their coolant material. The use of Pb-Bi coolant gave a low burnup reactivity swing and negative or less positive coolant void coefficient with harder neutron spectrum. By a lower burnup reactivity swing and higher conversion ratio of Pb-Bi cooled core, the total amount of TRU consumption was found to be small compared with Na cooled core despite of the higher MA consumption ratio of Pb-Bi cooled core. However, Pb-Bi cooled reactor have a lager margin in the coolant void coefficient, so that a variable MA composition can be loaded in the core. Accordingly, even though the Pb-Bi cooled TRU burner has not effectiveness on TRU burning in the same geometry and material condition, a flexible MA loading is envisaged to result in 10 times larger MA burning amount, still preserving a low coolant void worth

  11. Atucha II: building a 745 MWe pressure-vessel PHWR in the Argentine

    International Nuclear Information System (INIS)

    Madero, C.C.

    1982-01-01

    The history of the development of nuclear power in Argentina is outlined. Future policy is described. The aim is to close the fuel cycle and secure a domestic capability to design and construct nuclear power plants. (U.K.)

  12. Demonstration of a 1 MWe biomass power plant at USMC Base Camp Lejeune

    International Nuclear Information System (INIS)

    Cleland, J.; Purvis, C.R.

    1997-01-01

    A biomass energy conversion project is being sponsored by the U.S. Environmental Protection Agency (EPA) to demonstrate an environmentally and economically sound electrical power option for government installations, industrial sites, rural cooperatives, small municipalities, and developing countries. Under a cooperative agreement with EPA, Research Triangle Institute is initiating operation of the Camp Lejeune Energy from Wood (CLEW) biomass plant. Wood gasification combined with internal combustion engines was chosen because of (1) recent improvements in gas cleaning, (2) simple, economical operation for units less than 10 MW, and (3) the option of a clean, cheap fuel for the many existing facilities generating expensive electricity from petroleum fuels with reciprocating engines. The plant incorporates a downdraft, moving bed gasifier utilizing hogged waste wood from the Marine Corps Base at Camp Lejeune, NC. A moving bed bulk wood dryer and both spark ignition and diesel engines are included. Unique process design features are briefly described relative to the gasifier, wood drying, tar separation, and process control. A test plan for process optimization and demonstration of reliability, economics, and environmental impact is outlined. (author)

  13. Experimental activities related to safety systems of PWR NPPs (900 MWe). Sump screen plugging issue

    International Nuclear Information System (INIS)

    Vicena, I.; Soltesz, B.; Batalik, J.; Gubco, V.; Liska, M.; Galusekova, D.; Klementova, A.; Mattei, J. M.; Armand, Y.

    2006-01-01

    In this presentation authors present the sump plugging issue - ability of the filtration systems to catch debris (generated during the LOCA accident) to maintain the recirculating water the CSS and the ECCS need. Assessment has been focused upon the CPY series 900 MW e (28 reactors). Typical involved debris is: - the insulation for all the plants is made of glass wool (45 kg/m 3 ). Glass-wool is contained in sheet metal panels using attachment mechanism; - paints; - concrete; -dust. Effect of water chemistry on debris and filtration process ELISA loop is described.

  14. Experimental studies of 350 Mw(e) heterogeneous LMFBR cores at ZPPR

    International Nuclear Information System (INIS)

    Collins, P.J.; McFarlane, H.F.; Beck, C.L.; Lineberry, M.J.; Carpenter, S.G.; Ducat, G.A.; Gasidlo, J.M.; Goin, R.W.

    1979-01-01

    The annular heterogeneous concept has been adopted for the Clinch River Breeder Reactor. To provide verification of design and safety analysis, fourteen core configurations were studied at ZPPR. The results and analysis of the early assemblies, ZPPR-7A, 7B, and 7C, have been previously reported and only the more significant conclusions are summarized here. Later assemblies, 7D to 7H, were designed to provide data on variations of control rod and internal blanket arrangements. Assemblies 8A to 8E provided data on the replacement of uranium by thorium in various blanket subassemblies

  15. Improvement of a feed and bleed process for a 900 MWe NPP using the SIMPACT simulator

    International Nuclear Information System (INIS)

    Pochard, R.; Jedrzejewski, F.; Mazauric, X.; Cartenon, P. Y.

    2000-01-01

    A sensitivity study related to the improvement of a feed and bleed process was carried out with the SIPACT simulator. The scenario analysed here is related to a total loss of feed water on a French 900MW NPP. In a previous study we were looking extensively to the effects of bleeding with the three relief valves and its time of initiation. In the new calculations, limited bleed, by opening only one or two relief valves, was initiated at the minimum of mass and at the time which corresponds to the beginning of the heat transfer degradation in the steam generators. The analysis of the results shows that the in-vessel mass and the safety were improved when the number of actuated relief valves was reduced. But on the contrary the pressure reduction was limited by the performance of the HPIS and the equilibrium with the outlet flow from the relief valves. From these results a scenario with the consecutive opening of the three relief valves so as to depressurise while at the same time trying to optimise the in-vessel mass balance was proposed with a possible automation. (author)

  16. Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

    A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

  17. Software/firmware design specification for 10-MWe solar-thermal central-receiver pilot plant

    Energy Technology Data Exchange (ETDEWEB)

    Ladewig, T.D.

    1981-03-01

    The software and firmware employed for the operation of the Barstow Solar Pilot Plant are completely described. The systems allow operator control of up to 2048 heliostats, and include the capability of operator-commanded control, graphic displays, status displays, alarm generation, system redundancy, and interfaces to the Operational Control System, the Data Acquisition System, and the Beam Characterization System. The requirements are decomposed into eleven software modules for execution in the Heliostat Array Controller computer, one firmware module for execution in the Heliostat Field Controller microprocessor, and one firmware module for execution in the Heliostat Controller microprocessor. The design of the modules to satisfy requirements, the interfaces between the computers, the software system structure, and the computers in which the software and firmware will execute are detailed. The testing sequence for validation of the software/firmware is described. (LEW)

  18. Compressors. These little things that improve the operation of air conditioners. Danfoss-Turbocor: magnetic bearings for a centrifugal compressor. Copeland: the group stresses on the Digital power variation; Dossier compresseurs. Ces petits plus qui ameliorent le fonctionnement des climatiseurs. Danfoss-Turbocor: des paliers magnetiques pour un compresseur centrifuge. Copeland: le groupe met l'accent sur la variation de puissance Digital

    Energy Technology Data Exchange (ETDEWEB)

    Nicolas, J.

    2005-09-01

    This dossier about compressors for air conditioners comprises three articles dealing with: the improvements made by manufacturers of air-conditioning systems to increase the coefficient of performance and the lifetime of compressors, to reduce the refrigerant leaks and to reduce the power consumption; the electromagnetic bearings, the speed variation and the double stage compression used in the Danfoss-Turbocor centrifugal compressor; and the 'Digital' mechanical power variation system used by Copeland which does not change the motor velocity nor the operation limits of the compressor. (J.S.)

  19. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 1: Executive summary

    Science.gov (United States)

    1981-01-01

    Main elements of the design are identified and explained, and the rationale behind them was reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates are presented, and the engineering issues that should be reexamined are identified. The latest (1980-1981) information from the MHD technology program is integrated with the elements of a conventional steam power electric generating plant.

  20. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER) supplement. Magnet system special investigations

    Science.gov (United States)

    1981-01-01

    The results of magnet system special investigations listed below are summarized: 4 Tesla Magnet Alternate Design Study; 6 Tesla Magnet Manufacturability Study. The conceptual design for a 4 Tesla superconducting magnet system for use with an alternate (supersonic) ETF power train is described, and estimated schedule and cost are identified. The magnet design is scaled from the ETF 6 T Tesla design. Results of a manufacturability study and a revised schedule and cost estimate for the ETF 6 T magnet are reported. Both investigations are extensions of the conceptual design of a 6 T magnet system performed earlier as a part of the overall MED-ETF conceptual design described in Conceptual Design Engineering Report (CDER) Vol. V, System Design Description (SDD) 503 dated September, 1981, DOE/NASA/0224-1; NASA CR-165/52.

  1. The 10 MWe solar thermal central receiver pilot plant solar facilities design integration, RADL item 1-10

    Science.gov (United States)

    1980-08-01

    Work on the plant support subsystems and engineering services is reported. The master control system, thermal storage subsystem, receiver unit, and the beam characterization system were reviewed. Progress in program management and system integration is highlighted.

  2. Design and construction of the PCPV for the 300 MWe THTR nuclear power station in West Germany

    International Nuclear Information System (INIS)

    Bremer, F.

    1976-01-01

    In July 1972 the order was placed in Germany for the first PCPV comprising concrete structure, liner, cooling system and insulation to a consortium under the direction of KRUPP UNIVERSALBAU. The prestressed concrete structure itself was designed and constructed by this company. Extensive tests were carried out on the limestone concrete to establish all the physical properties. Special efforts were made to produce a mix which was both pumpable and generated a minimum amount of heat of hydration. As a departure from normal practice, the cylindrical parts of the vessel are constructed in complete rings up to 2m in height and of the full wall thickness. Experiments showed that, for this method of construction, the temperature difference between the old and the new concrete should not be allowed to exceed 5 0 C. To achieve this, ice cooled water is used in the concrete mix and, in the summer time, liquid nitrogen is added at the time of mixing. The thermal behavior of the concrete has been monitored throughout the construction period. A novel construction feature worth mentioning is that the internal insulation and parts of the core structure were already erected before the construction of the concrete cylinder was complete. This was achieved by providing a temporary closure at the top of the cylinder to maintain clean conditions below. The overall stress calculations and the detailed stress pattern for the lower half of the vessel were carried out by using an axi-symmetric computer program but, for the upper half of the cylinder, a three-dimensional analysis was necessary (due to its geometric arrangement). To prove the safety of the vessel a structural model was used from which the mode of failure was found using a kinematic chain and thus the factor of safety established. A secondary line of safety is the integrity of the liner. (author)

  3. Environmental effects of solar thermal power systems: ecological observations during construction of the Barstow 10 MWe pilot STPS

    Energy Technology Data Exchange (ETDEWEB)

    Turner, F.B. (ed.)

    1981-10-01

    The environmental monitoring plan used consists of comparisons of a few meteorological variables and changes in the states of a limited array of indicator species or assemblages of species of plants and animals. Observations inlude aerial photography of the site, saltation meter measurements downwind from the site to measure fluxes of windblown sand, measurements of airborne particulates and atmospheric pollutants, and baseline temperature profiles made at two sites near the heliostat field to measure micro-meteorological patterns. Observations were made of annual plants both in off-field plots and in heliostat field, of shrubs, birds, rodents, reptiles, and sensitive species listed as rare or endangered. (LEW)

  4. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  5. Some safety related characteristics of Phenix, a 250 MWe fast reactor -1989 and 1990 negative reactivity trip investigations

    International Nuclear Information System (INIS)

    Chaumont, J.M.; Goux, D.; Martin, L.

    1993-01-01

    The main characteristics of the Phenix core control are summarized. The current state of the investigations related to the 1989 and 1990 negative reactivity transients are presented with emphasis on the results of the very low power tests recently performed. (authors). 5 figs., 2 refs

  6. Numerical investigation on the flow, combustion, and NOX emission characteristics in a 660 MWe tangential firing ultra-supercritical boiler

    Directory of Open Access Journals (Sweden)

    Wenjing Sun

    2016-02-01

    Full Text Available A three-dimensional numerical simulation was carried out to study the pulverized-coal combustion process in a tangentially fired ultra-supercritical boiler. The realizable k-ε model for gas coupled with discrete phase model for coal particles, P-1 radiation model for radiation, two-competing-rates model for devolatilization, and kinetics/diffusion-limited model for combustion process are considered. The characteristics of the flow field, particle motion, temperature distribution, species components, and NOx emissions were numerically investigated. The good agreement of the measurements and predictions implies that the applied simulation models are appropriate for modeling commercial-scale coal boilers. It is found that an ideal turbulent flow and particle trajectory can be observed in this unconventional pulverized-coal furnace. With the application of over-fire air and additional air, lean-oxygen combustion takes place near the burner sets region and higher temperature at furnace exit is acquired for better heat transfer. Within the limits of secondary air, more steady combustion process is achieved as well as the reduction of NOx. Furthermore, the influences of the secondary air, over-fire air, and additional air on the NOx emissions are obtained. The numerical results reveal that NOx formation attenuates with the decrease in the secondary air ratio (γ2nd and the ratio of the additional air to the over-fire air (γAA/γOFA was within the limits.

  7. Investigation of techniques for the application of safeguards to the CANDU 600 MW(e) nuclear generating station

    International Nuclear Information System (INIS)

    Smythe, W.D.

    1978-06-01

    A cooperative program with the Canadian Atomic Energy Control Board, Atomic Energy of Canada Limited and the IAEA was established in 1975: to determine the diversion possibilities at the CANDU type reactors using a diversion path analysis; to detect the diversion of nuclear materials using material accountancy and surveillance/containment. Specific techniques and instrumentation, some of which are unique to the CANDU reactor, were developed. 10 appendices bring together the relevant reports and memoranda of results for the Douglas Point Program

  8. Development of innovative tools based on fuelling machine for ageing management of coolant channels of 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Dev, Mahender; Roy, Shyamal; Bhattachrya, Sambit; Singh, Jit Pal; Patel, R.J.; Agarwal, R.G.

    2006-01-01

    PHWR coolant channels are required to be inspected periodically to satisfy the regulatory requirement, to provide information about known or suspected problem and to provide information to assist in future design. This paper describes these tools and techniques, their capabilities and experience of implementing these in reactor site

  9. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  10. Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-16

    The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

  11. Deposit Formation in a 150 MWe Utility PF-Boiler during Co-combustion of Coal and Straw

    DEFF Research Database (Denmark)

    Andersen, Karin Hedebo; Frandsen, Flemming; Hansen, P. F. B.

    2000-01-01

    A conventional pc-fired boiler at the Danish energy company I/S Midtkraft has been converted to coal-straw co-combustion, and a 2 year demonstration program was initiated in January 1996, addressing several aspects of coal-straw co-combustion. Deposition trials were performed as part of the demon......A conventional pc-fired boiler at the Danish energy company I/S Midtkraft has been converted to coal-straw co-combustion, and a 2 year demonstration program was initiated in January 1996, addressing several aspects of coal-straw co-combustion. Deposition trials were performed as part...... problematic deposits. Go-firing straw also caused a change in the structure of the upstream deposits. During coal combustion an ordered, "finger" structure of the larger particles with small particles between was observed, whereas during co-combustion a more random deposition of the larger particles among...... arise when burning other coals, particularly coals with a high S or alkali metal content or a low content of ash. The behavior of K, Ca, S, and Cl was evaluated by use of thermodynamic calculations. The thermodynamically stable species agree with the observed behavior in the experiments, i.e. formation...

  12. 10-MWe solar-thermal central-receiver pilot plant: collector subsystem foundation construction. Revision No. 1

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-18

    Bid documents are provided for the construction of the collector subsystem foundation of the Barstow Solar Pilot Plant, including invitation to bid, bid form, representations and certifications, construction contract, and labor standards provisions of the Davis-Bacon Act. Instructions to bidders, general provisions and general conditions are included. Technical specifications are provided for the construction. (LEW)

  13. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    Energy Technology Data Exchange (ETDEWEB)

    Madhuresh, R; Nagarajan, R; Jit, I; Sanatkumar, A [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like `stay-put`, `gravity- fill`, `D{sub 2}0- steaming` etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs.

  14. THESEUS - the first utility-scale 50 MWe. THErmal Solar EUropean power Station for the island of Crete, Greece

    Energy Technology Data Exchange (ETDEWEB)

    Cobi, A. [PreussenElektra AG, Hannover (Germany); Tzatzanis, A.

    1997-12-31

    When the European Commission`s Directorate for Energy (DG XVII) opened the second application window for THERMIE proposals under the Fourth Framework Programme on September 15, 1995, Europe`s solar thermal power community discovered a remarkable modification to the previous call for proposals. For the first time since solar thermal electricity was introduced for THERMIE demonstration projects, it now became eligible for EU financial support. Right now, the European Commission committed funding for the design and engineering phase of the THESEUS project implementation. (orig.)

  15. Improvement of two calculation methods of the detectors activation and on the PWR 900 MWe vessel. Comparison with the experiment

    International Nuclear Information System (INIS)

    Kitsos, S.

    1992-11-01

    The conditions of the vessel ageing (damages through irradiation) that mostly determine the life time of a nuclear reactor depend on the dose received. For the determination of this dose we use two calculation methods: one exact method using the TRIPOLI code that solves the Boltzmann equation with the Monte-Carlo method and one simplified method based on the point-kernel method. The advantages of the second method which is fast (easy reproduction of the results) and deterministic (the effects of difference are possible) compared to the first one which is long and statistical make its development necessary. The qualifications of these methods are done by comparison with the experiment that we reach as following: for the first method, we check the programming of TRIPOLI code (representativity of the collision) and its alignment with SN codes, and we modify the second method in order to use variable linear attenuation coefficients inside each medium to represent better the effects of the spectrum and of the reflection. As part of the checking of the basic physical data and their mode of representation, we present a study of the influence of the energy group averaging of the cross sections and of the number of the groups, as well as study of the influence of the cross sections origin

  16. The application of redundancy-related basic safety principles to the 1400 MWE reactor core standby cooling system

    International Nuclear Information System (INIS)

    Bertrand, R.

    1990-01-01

    This memorandum shall provide the background for the work of the European Community Commission which is to analyze safety principles relating to redundancy. The redundancy-related basic safety principles applied in French nuclear power plants are the following: . the single-failure criterion, . provisions additional to application of the single-failure criterion. These are mainly provisions made at the design stage to minimize risks associated with common cause failures or the risks of human error which can lead to such failures: - protection against hazards of internal and external origin, - the geographical or physical separation of equipment, - the independence of electrical power supplies and distribution systems, - the additional resources and associated operating procedures making it possible to accommodate total loss of the safety systems. The scope also includes the operating rules which ensure availability of redundant safety-related equipment. The provisions relating to the single-failure criterion are detailed in Basic Safety Rule 1.3.A appended. The application of these principles proposed by the operating organization and accepted by the safety authorities for the design and operation of the standby core cooling system (System RIS) is explained

  17. Refitting of the 'Celimene' hot cell for following up the fuel assembly of 900 MWe PWR power reactors

    International Nuclear Information System (INIS)

    Lhermenier, Andre; Van Craeynest, J.-C.

    1980-05-01

    The 'Celimene' cell adjoining the EL3 reactor provides for the acceptance, handling and the examination of irradiated fuel assemblies from power reactors (length approximately 4m, weight approximately 700 kg). Within the framework of the PWR fuel behavior follow-up or reprocessing, it is possible to extract an assembly representative of the normal fuel cycle, carry out non destructive tests on this assembly, extract pencils from it and re-insert this assembly, after refitting the head, into the normal fuel cycle for handling in a reprocessing plant or storage pond. Given suitable refitting techniques, the re-irradiation of the assembly can be considered after examination. Significant changes have been made to the buildings and the hoist facilities for handling very heavy flasks. It was necessary to rearrange the handling, machining and in-cell storage facilities. The development of an inspection rig will make it possible, some time in 1980, to carry out non destructive tests of assemblies, optical and metrological examination of assemblies prior to dismantling or of the structure after dismantling [fr

  18. Full Scale Deposition Trials at 150 MWe PF-boiler Co-firing COal and Straw: Summary of Results

    DEFF Research Database (Denmark)

    Andersen, Karin Hedebo; Frandsen, Flemming; Hansen, Peter Farkas Binderup

    1999-01-01

    A conventional PF-fired boiler at the Danish energy company I/S Midtkraft has been converted to coal-straw co-combustion and a two-year demonstration programme was initiated in January 1996 addressing several aspects of coal-straw co-combustion. Deposition trials were performed as part...... during co-combustion with straw. In addition, where Fe dominated upstream deposits are found in the hottest positions during pure coal combustion, Ca, and to some degree Si, are playing the major role during co-combustion. The addition of straw to the fuel is also seen to lead to a change in the texture...... of the upstream deposits, from an ordered dendritic structure of the larger particles with small particles in between during pure coal combustion, to a more random deposition of the larger particles among the small during co-combustion. No deposition of chlorine species was observed in the SEM-EDX analysis...

  19. Design study of a PWR of 1300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, J.G.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and of the thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. Reactor core design calculations were performed for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM-2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. No technical problem that precluded the utilization of any of the options analyzed was found. The savings in uranium ore introduced by the thorium cycle with fuel recycling ranges from 13% to 52% as compared with the usual uranium once-through cycle; the SWU savings goes from 13% to 22%. (Author) [pt

  20. Opinion on the demonstration of the 900 MWe reactor vessels in-service behaviour after their third decennial inspection

    International Nuclear Information System (INIS)

    2010-01-01

    In this report, an expert group comments and assesses how sufficient are the demonstration and the actions performed by EDF to justify the in-service behaviour of nuclear reactor vessels. More precisely, it comments and discusses the different steps of the EDF demonstration: follow-up of the fluence received by the vessels, identification of the most severe transients and thermodynamic calculations, behaviour of irradiated materials, mechanical analysis, in-service control and follow-up plan, ageing management. Recommendations are then formulated

  1. Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

    International Nuclear Information System (INIS)

    Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

    1996-01-01

    Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

  2. The 10 MWe Solar Thermal Central Receiver Pilot Plant: Solar facilities design integration. Pilot-plant station manual (RADL Item 2-1). Volume 1: System description

    Science.gov (United States)

    1982-09-01

    The complete Barstow Solar Pilot Plant is described. The plant requirements and general description are presented, the mechanical, electric power, and control and instrumentation systems as well as civil engineering and structural aspects and the station buildings are described. Included in the mechanical systems are the heliostats, receiver, thermal storage system, beam characterization system, steam, water, nitrogen, and compressed air systems, chemical feed system, fire protection system, drains, sumps and the waste disposal systems, and heating, ventilating, and air conditioning systems.

  3. Receiver subsystem analysis report (RADL Item 4-1). The 10-MWe solar thermal central-receiver pilot plant: Solar-facilities design integration

    Science.gov (United States)

    1982-04-01

    The results of thermal hydraulic, design for the stress analyses which are required to demonstrate that the receiver design for the Barstow Solar Pilot Plant satisfies the general design and performance requirements during the plant's design life are presented. Recommendations are made for receiver operation. The analyses are limited to receiver subsystem major structural parts (primary tower, receiver unit core support structure), pressure parts (absorber panels, feedwater, condensate and steam piping/components, flash tank, and steam mainfold) and shielding.

  4. A 10-MWe solar-thermal central-receiver pilot plant: Solar facilities design integration. Plant operating/training manual (RADL-Item 2-36)

    Science.gov (United States)

    1982-07-01

    Plant and system level operating instructions are provided for the Barstow Solar Pilot Plant. Individual status instructions are given that identify plant conditions, process controller responsibilities, process conditions and control accuracies, operating envelopes, and operator cautions appropriate to the operating condition. Transition operating instructions identify the sequence of activities to be carried out to accomplish the indicated transition. Most transitions involve the startup or shutdown of an individual flowpath. Background information is provided on collector field operations, and the heliostat groupings and specific commands used in support receiver startup are defined.

  5. Industrial Application of an Improved Multiple Injection and Multiple Staging Combustion Technology in a 600 MWe Supercritical Down-Fired Boiler.

    Science.gov (United States)

    Song, Minhang; Zeng, Lingyan; Chen, Zhichao; Li, Zhengqi; Zhu, Qunyi; Kuang, Min

    2016-02-02

    To solve the water wall overheating in lower furnace, and further reduce NOx emissions and carbon in fly ash, continuous improvement of the previously proposed multiple injection and multiple staging combustion (MIMSC) technology lies on three aspects: (1) along the furnace arch breadth, changing the previously centralized 12 burner groups into a more uniform pattern with 24 burners; (2) increasing the mass ratio of pulverized coal in fuel-rich flow to that in fuel-lean flow from 6:4 to 9:1; (3) reducing the arch-air momentum by 23% and increasing the tertiary-air momentum by 24%. Industrial-size measurements (i.e., adjusting overfire air (OFA) damper opening of 20-70%) uncovered that, compared with the prior MIMSC technology, the ignition distance of fuel-rich coal/air flow shortened by around 1 m. The gas temperature in the lower furnace was symmetric and higher, the flame kernel moved upward and therefore made the temperature in near-wall region of furnace hopper decrease by about 400 °C, the water wall overheating disappeared completely. Under the optimal OFA damper opening (i.e, 55%), NOx emissions and carbon in fly ash attained levels of 589 mg/m(3) at 6% O2 and 6.18%, respectively, achieving NOx and carbon in fly ash significant reduction by 33% and 37%, respectively.

  6. Hollow fiber membrane contactors for CO2 capture: modeling and up-scaling to CO2 capture for an 800 MWe coal power station

    NARCIS (Netherlands)

    Kimball, E.; Al-Azki, A.; Gomez, A.; Goetheer, E.L.V.; Booth, N.; Adams, D.; Ferre, D.

    2014-01-01

    A techno-economic analysis was completed to compare the use of Hollow Fiber Membrane Modules (HFMM) with the more conventional structured packing columns as the absorber in amine-based CO2capture systems for power plants. In order to simulate the operation of industrial scale HFMMsystems, a

  7. On-site A.C. electric power sources for 900 MWe French nuclear power reactors: reliability and importance for safety

    Energy Technology Data Exchange (ETDEWEB)

    Milhem, J. L.; Gros, G. [Commissariat a l' Energie Atomique, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

    1986-02-15

    After presenting briefly the new provisions laid down by the Electricite de France to meet a total electrical power loss, the main elements of the probabilistic study concerning the corresponding risk described: reliability data of internal sources used, results of risk Improvement brought by the new measures, importance for Internal source before and after Implementation of the new measures. (authors)

  8. Probabilistic determination of operating rules for 6.6 kV power supplies to 900 MW(e) pressurized-water units

    International Nuclear Information System (INIS)

    Blin, A.; Carnino, A.; Boursier, M.; Greppo, J.F.

    1975-01-01

    The 6.6 kV power supplies to the safety systems of the 900 MW PWR units being built by Electricite de France (EDF) are ensured by four redundant sources within or outside the site. Because of this redundancy, non-availability of one or more of the sources does not jeopardize the electricity supply - it merely means that the unit must be shut down deliberately for safety reasons at some time. The proposed method is a probabilistic approach which makes it possible to determine how long the unit can reasonably continue on power in such circumstances. The study described in the paper is based on operating results recorded by EDF for equipment similar to that envisaged for future nuclear power stations and in service in existing ones. The amount of information gathered and the method employed for gathering it permit a realistic assessment of the reliability parameters of the equipment (failure and repair rates). To allow for the fact that the equipment can be repaired, the method proper involves use of the Markov model, with which one can find, for each configuration of the system, the change over time of the probability P 0 of a simultaneous failure of all power sources. Use of this method requires prior conversion of the actual concept to an equivalent concept. The sensitivity of P 0 to the parameters is studied for each case, a reasonable uncertainty range being obtained for P 0 . Common mode failures for external or internal sources are introduced in parametric form so that the parameter value beyond which they can be neglected is determined. On the basis of the results of the study, the authors propose the adoption of a comparative criterion whereby any mode of operation in degraded conditions (one or more sources not available) is permitted provided that the corresponding value of P 0 is always lower than the maximum value attained by it in the reference situation (all sources available). Applying this criterion, one can determine with the help of graphs the operating periods which must not be exceeded. Thus, in the case of non-availability of an external source (the grid) one obtains a maximum period of 12 to 36 hours; in the case of non-availability of an internal source (diesel generator) the maximum period is 150 to 240 hours

  9. On-site A.C. electric power sources for 900 MWe french nuclear power reactors: reliability and importances for safety

    International Nuclear Information System (INIS)

    Milhem, J.L.; Gros, G.

    1985-10-01

    After presenting briefly the new provisions laid down by the Electricite de France to meet a total electrical power loss, the main elements of the probabilistic study concerning the corresponding risk are described: reliability data of internal sources used, results of risk improvement brought by the new measures, importance for internal source before and after implementation of the new measures

  10. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  11. Assessment of generic solar thermal systems for large power applications: analysis of electric power generating costs for systems larger than 10 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Apley, W.J.; Bird, S.P.; Brown, D.R.; Drost, M.K.; Fort, J.A.; Garrett-Price, B.A.; Patton, W.P.; Williams, T.A.

    1980-11-01

    Seven generic types of collectors, together with associated subsystems for electric power generation, were considered. The collectors can be classified into three categories: (1) two-axis tracking (with compound-curvature reflecting surfaces); (2) one-axis tracking (with single-curvature reflecting surfaces); and (3) nontracking (with low-concentration reflecting surfaces). All seven collectors were analyzed in conceptual system configurations with Rankine-cycle engines. In addition, two of the collectors were analyzed with Brayton-cycle engines, and one was analyzed with a Stirling-cycle engine. With these engine options, and the consideration of both thermal and electrical storage for the Brayton-cycle central receiver, 11 systems were formulated for analysis. Conceptual designs developed for the 11 systems were based on common assumptions of available technology in the 1990 to 2000 time frame. No attempt was made to perform a detailed optimization of each conceptual design. Rather, designs best suited for a comparative evaluation of the concepts were formulated. Costs were estimated on the basis of identical assumptions, ground rules, methodologies, and unit costs of materials and labor applied uniformly to all of the concepts. The computer code SOLSTEP was used to analyze the thermodynamic performance characteristics and energy costs of the 11 concepts. Year-long simulations were performed using meteorological and insolation data for Barstow, California. Results for each concept include levelized energy costs and capacity factors for various combinations of storage capacity and collector field size.

  12. Commissioning of the THTR-300-MWe prototype power plant - A milestone for further application of this high-temperature reactor line

    International Nuclear Information System (INIS)

    Simon, M.; Baust, E.; Schoening, J.

    1986-10-01

    With the completion of the THTR 300 and the development of the follow-on plant HTR 500, the BBC/HRB company group has taken the pebble bed high-temperature reactor to the threshold of the commercial stage. The HTR is an important innovation in the field of reactor technology which can play an important role in the intermediate and long-term supply of safe, environmental friendly and economic energy. The power level of 550 MW meets the requirements of the present energy market which shows a trend towards smaller power units as a result of grid size, investment effort, and the slower increase in electricity demand in industrial nations. The advantages of the high-temperature reactor, such as high thermal efficiency, low waste heat, low radiation exposure of operating and maintenance personnel, high inherent safety, simple mode of operation, flexible fuel cycle with the potential to extend fuel resources, high availability, are currently uncontested and will represent the future standards for the peaceful uses of nuclear energy. For special applications in industry (steam and electric power as a cogeneration product) and in case of special siting conditions (near industrial centers), BBC/HRB developed a small 100 MW HTR, which can also be constructed as a 200 MW twin plant at favorable cost conditions. For an economic use of domestic coal in a processed form, the HTR represents the optimum solution as to economic and environmental aspects as well as extension of resources, especially if combined with conventional gasification procedures and in direct application of nuclear process heat at high gas temperatures of about 950 deg. C. In this field the development of the heat-exchanging components remains to be completed, before commercial application will be possible. The HTR is particularly well suited for erection in developing countries and industrial threshold countries which turn to nuclear energy for the first time. On an international level the interest in the pebble bed high-temperature reactor has also increased recently. Thus the HTR is of great importance to electric power industry and industrial development

  13. Investigations on construction material and construction concepts in order to obtain dose-reducing effects in the dismantling of the biological shield of a 1300 MWe-PWR

    International Nuclear Information System (INIS)

    Bittner, A.; Jungwirth, D.; Knell, M.; Schnitzler, L.

    1984-04-01

    Numerical values of neutron fluxes, activations, dose rates etc. as a function of characteristic values of materials required for optimization purposes to reduce the radiation effect of the biological shield of a PWR are not available. Design concepts are presented for biological shields of PWRs made of concrete with respect to both the most suitable application of materials and the design principles aiming at reduced radiation exposure as compared to present designs during entering, waste disposal and ultimate storage. To evaluate the present-state design the above values have been calculated. Suggested alternative designs are biological shields with selective material application, built from precast elements with or without boron carbide layer arranged in front of it. (orig./HP) [de

  14. Results of performance and emission testing when co-firing blends of dRDF/COAL in a 440 MWe cyclone fired combustor

    International Nuclear Information System (INIS)

    Ohlsson, O.O.

    1993-01-01

    Argonne National Laboratory (ANL) together with the University of North Texas (UNT) have developed an improved method for converting refuse (residential, commercial and institutional waste) into an environmentally safe and economical fuel. In this method, recyclable metals, glass, and some plastic products are separated from the refuse. The remaining fraction, consisting primarily of cellulosic materials is then combined with a calcium hydroxide binding additive and formed into cylindrical pellets. These pellets are dense and odorless, can be stored for extended periods of time without biological or chemical degradation, and due to their increased bulk density are more durable and can be more easily conveyed, handled, and transported than other types of waste-derived fuel pellets. Laboratory and pilot-scale research studies, followed by full-scale combustion tests undertaken by DOE, ANL and UNT, in June--July of 1987 have indicated that binder-enhanced dRDF pellets can be successfully cofired with high sulfur coal in spreader-stoker combustors. The results of these combustion tests indicated significant reductions of SO 2 , NO x and CO 2 in the flue gases, and the reduction of heavy metals and organics in the ash residue. Dioxins and furans, both in the flue gas and in the ash residues were below detectable levels. Additional commercial-scale combustion tests have recently been conducted by DOE, NREL, ANL and several industrial participants including Otter Tail Power Company, Reuter, Inc., XL Recycling and Marblehead Lime Company, under a collaborative research and development agreement (CRADA). A large 440 MW e cyclone-fired combustor was tested at Big Stone City, South Dakota on October 26--27, 1992. This paper describes the cyclone-fired combustion tests, the flue gas emission and ash samples that were collected, the analyses that were performed on these samples, and the final test results

  15. Numerical study of flow, combustion and emissions characteristics in a 625 MWe tangentially fired boiler with composition of coal 70% LRC and 30% MRC

    Science.gov (United States)

    Sa'adiyah, Devy; Bangga, Galih; Widodo, Wawan; Ikhwan, Nur

    2017-08-01

    Tangential fired boiler is one of the methods that can produce more complete combustion. This method applied in Suralaya Power Plant, Indonesia. However, the boiler where supposed to use low rank coal (LRC), but at a given time must be mixed with medium rank coal (MRC) from another unit because of lack of LRC coal. Accordingly to the situation, the study about choosing the right position of LRC and MRC in the burner elevation must be investigated. The composition of coal is 70%LRC / 30%MRC where MRC will be placed at the lower (A & C - Case I)) or higher (E & G - Case II) elevation as the cases in this study. The study is carried out using Computational Fluid Dynamics (CFD) method. The simulation with original case (100%LRC) has a good agreement with the measurement data. As the results, MRC is more recommended at the burner elevation A & C rather than burner elevation E & G because it has closer temperature (880 K) compared with 100%LRC and has smaller local heating area between upper side wall and front wall with the range of temperature 1900 - 2000 K. For emissions, case I has smaller NOx and higher CO2 with 104 ppm and 15,6%. Moreover, it has samller O2 residue with 5,8% due to more complete combustion.

  16. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

    The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

  17. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Energy Technology Data Exchange (ETDEWEB)

    Gourguechon, B.

    1993-12-31

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  18. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  19. Seismic qualification of moderator system pump-motor units for RAPP-3,4 and KAIGA-1,2 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Neelwarne, A.; Soni, R.S.; Kushawaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1992-01-01

    Smooth operation of active components like primary heat transport pumps, moderator pumps, emergency core cooling pumps etc. is always required to ensure safety of any nuclear power plants in case of normal as well as abnormal conditions such as earthquake loading. In order to ensure the functional requirement of such rotating equipment, is necessary to demonstrate, either through theoretical means or through experimental means, that in an event like earthquake loading, the static parts and the rotating parts of the equipment do not rub against each other giving rise to trouble during their operation. The moderator system pump units for RAPP-3,4 and Kaiga-1,2 have been analysed theoretically to demonstrate the structural integrity of various components of the unit as well as the functional requirement during an earthquake loading. A detailed Finite Element Model (FEM) was prepared for this which includes the modelling of static parts, rotating parts, anti-friction bearings and fluid-film journal bearings. Response spectrum analysis of the unit was carried out using the applicable floor response spectra for RAPP-3,4 and Kaiga-1,2 sites. It was concluded from this analysis that the pump-motor unit analysed meets the required design intent in terms of structural integrity and operability of the unit. The present report gives a detailed description of the problem, the development of FEM model, results and the conclusions arrived at. (author). 23 refs., 9 tabs., 17 figs

  20. Problems in Bearings and Lubrication

    Science.gov (United States)

    1982-08-01

    tant verticale qulhorizontale. Le palier niagn~tique actif du type axial fonctionne selon le principe de ferro-attraction. La force d’origlne filec...contractions. * REFERENCES III D. Dini, G. Nardi and G. Pizzolante Leuzzi, " Turbina in circuito chiuso per espaimeione di gas idrogeno fino a temperatura

  1. Protéger les mangroves du Cambodge | CRDI - Centre de ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    CRDI), un groupe de chercheurs du palier national et de la province du Koh Kong utilise une nouvelle approche pour gérer les ressources de la mangrove. Cette approche, qui fait appel à l'esprit d'initiative des membres de la ...

  2. Vers l'integration des textes dans le programme du fran9ais langue ...

    African Journals Online (AJOL)

    Au Malawi, on enseigne le fran9ais en tant que langue etrangere depuis une quarantaine d'annees .... !'organisation institutionnelle du systeme educatif du Malawi, le palier secondaire en ... Les objectifs du type : rediger une lettre, decrire une .... nouveau (rheme). Charolles souligne qu'il s'agit d'une performance delicate.

  3. Dynamic containment of gaseous effluents in the auxiliary buildings and reinjection of liquid effluents from these buildings back into the reactor building for 900 MWe PWRs under accident condition

    International Nuclear Information System (INIS)

    Demoulin, F.; Collinet, J.; Nguyen, C.

    1987-04-01

    Examination of the lessons to be learned from the accident of the Three Mile Island nuclear power plant on 20 March 1979 led the French Safety Authorities and EDF (Electricite de France) to adopt a series of measures intended to improve the performance of the containment of French PWRs, especially in the event of accident. Among the measures adopted, two of them contribute to the upgrading of the containment of nuclear island buildings, by reducing radioactivity constraints inside these buildings and by limiting radioactive releases into the environment. These are: (1) dynamic containment of auxiliary buildings likely to be contaminated following an accident, (2) reinjection back into the reactor building of liquid effluents arising in the auxiliary buildings. In this paper we shall discuss, for each measure, the approach to the problem and describe the arrangements made to arrive at a satisfactory solution [fr

  4. Electronuclear reactors - EDF. Synthesis of study appraisal and of modifications associated with the safety re-examination of 1300 MWe reactors after 30 years of operation (VD3 1300)

    International Nuclear Information System (INIS)

    Bigot, Franck

    2014-01-01

    This report states the opinion of the IRSN on the adequacy of safety improvements adopted by EDF and of the update of the safety demonstration, as well as their consistency with guidelines defined for this safety re-examination; on the modalities defined by EDF to assess the compliance and status of installations (topics related to the ageing of containment enclosure and vessels being excluded); on the adequacy of the validation performed by EDF in terms of organisational and human factors for the whole set of modifications related to reactor ageing management and obsolescence management (notably those related to the modernisation of control rooms); on the acceptability of modifications declared by EDF in April 2014 on issues related to design, implementation and exploitation with respect to elements of the Code of the Environment. Thus, the report addresses: the safety demonstration update and adopted improvements (studies of operating conditions and their radiological consequences with respect to the different identified risks, design of systems which are important for safety and civil works components, aggressions with an external or internal origin, probabilistic safety studies), the installation compliance and status (decennial tests, program of additional investigations), the modernisation of the control room and social, organisational and human aspects, and equipment modifications. A table in appendix indicates the different topics of the VD 1300 safety re-examination, the associated published IRSN opinions and reports, and the associated ASN letters. Another appendix contains a set of recommendations regarding rules, methods and accident studies for the Safety Report, for the primary circuit dilution risks, for the radiological consequences of severe accidents, for the safety of the nuclear fuel stored in a pool, for the modernisation of the control room and the organisational and human factors, for equipment modifications, and for different parts of the General Rules of Exploitation (RGE)

  5. Vers une (r)évolution du renseignement belge : la nécessaire émergence d'une communauté du renseignement

    OpenAIRE

    Leroy, Patrick

    2017-01-01

    Le renseignement belge entre dans une période de (r)évolution amenée par la crise des attentats qui secouent le sol européen. La tentations est grande pour les décideurs politiques de palier les "failles" du renseignement par des mesures radicales qui pourraient atteindre l'ADN, le coeur de métier du renseignement.

  6. Telecommunications and Universal Service : International ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Les services fournis au moyen de réseaux de télécommunications peuvent offrir des possibilités énormes dans les domaines suivants : éducation, soins de santé, production de revenu et accès à tous les paliers de gouvernement et communication avec eux. Pour en bénéficier, les pays doivent trouver des moyens d'offrir ...

  7. Mechanical pumps for liquid metals; Pompes mecaniques pour metaux liquides

    Energy Technology Data Exchange (ETDEWEB)

    Baumier, J; Gollion, H J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    The pumping of liquid metals by centrifugal pumps poses two principal problems. These are hermetic sealing of the rotating shaft and, its guidance where immersed in liquid metal. The solutions to the problems used on 13 experimental pumps are given here. The resolution of the guidance problem consists in the majority of cases in the utilisation of hydrostatic bearings. Accordingly, a theoretical study was instituted for the first time to calculate the bearings of the earlier pumps. After this, an experimental study was carried out, to check the theory by water tests. A relation for bearing calculation of pumps with diffusers is proposed. Finally the influence of the bearing elasticity on the shafts critical speed is studied. (authors) [French] Le pompage des metaux liquides, par des pompes centrifuges, pose 2 principaux problemes, qui sont: d'une part, la realisation d'une excellente etancheite au passage de l'arbre, d'autre part, son guidage sur la partie immergee dans le metal liquide. Les solutions retenues pour resoudre ces problemes sur 13 pompes experimentees sont presentees. Le probleme du guidage de l'arbre, a dans la majorite des cas ete resolu en utilisant un palier hydrostatique, aussi l'etude en a d'abord ete approfondie de facon theorique pour calculer les paliers des premieres pompes, puis experimentale pour controler la theorie, en effectuant des essais a l'eau. On propose une relation pour calculer les paliers des pompes a diffuseurs. On a en outre effectue une etude de l'influence de l'elasticite du palier hydrostatique sur la vitesse critique de l'arbre. (auteurs)

  8. La politique de financement public du sport au Cameroun. Quels enjeux pour quelle politique fiscale ?

    Directory of Open Access Journals (Sweden)

    Obam Richard Evina

    2016-01-01

    Pour ce faire, cinq paliers doivent être successivement franchis. Le premier serait constitué par l'adoption d'une loi sur l'organisation et la promotion des activités physiques et sportives au Cameroun ayant comme caractéristiques essentielles l'obligation de constitution des clubs en société anonyme à objet sportif (SAOS avant toute affiliation à une Fédération et la création d'une Direction Nationale de Contrôle et de Gestion (DNCG ; Le deuxième palier consisterait dans l'opérationnalisation de la loi de 2006 sur le partenariat public-privé qui permettrait aux entreprises privées d'intervenir dans le domaine de la construction des infrastructures sur le modèle Build-operate-transfer (BOT ; Le troisième palier serait l'adoption d'un dispositif fiscal incitatif pour le mécénat, le sponsoring, le bénévolat, le marchandising et les droits télé ; Enfin, outre la dynamisation de la coopération internationale dans la perspective de diversification des financements viendrait parachever cette architecture de reconstruction du mouvement sportif camerounais.

  9. SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

    International Nuclear Information System (INIS)

    Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.

    2003-01-01

    The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients

  10. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  11. Development and construction of nuclear power and nuclear heating stations in the USSR

    International Nuclear Information System (INIS)

    Schmidt, G.; Kirmse, B.

    1983-01-01

    The state-of-the-art of nuclear power technology in the USSR is reviewed by presenting characteristic data on design and construction. The review takes into consideration the following types of facilities: Nuclear power stations with 1000 MWe pressurized water reactors, with 1000 MWe pressure tube boiling water reactors, and with 600 MWe fast breeder reactors; nuclear heating power stations with 1000 MWe reactors and nuclear heating stations with 500 MWth boiling water reactors

  12. Shielding design for PWR in France

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  13. Nitrogen speciation in FGD waste water

    Energy Technology Data Exchange (ETDEWEB)

    Fogh, F. [Elsam A/S, Skaerbaekvaerket, Fredericia (Denmark); Smitshuysen, E.F. [Elsam A/S, Esbjergvaerket, Esbjerg (Denmark)

    2003-07-01

    Elsam operates six flue gas desulphurisation (FGD) units (2590 MWe): three wet FGD units (1440 MWe) and three semi-dry FGD units (1150 MWe). The paper presents the results of Elsam investigations covering nitrogen analysis of selected aqueous and solid streams together with nitrogen source and sink considerations in wet and semi-dry FGD plants. (orig.)

  14. Shippingport: A relevant decommissioning project

    International Nuclear Information System (INIS)

    Crimi, F.P.

    1988-01-01

    Because of Shippingport's low electrical power rating (72 MWe), there has been some misunderstanding on the relevancy of the Shippingport Station Decommissioning Project (SSDP) to a modern 1175 MWe commercial pressurized water reactor (PWR) power station. This paper provides a comparison of the major components of the reactor plant of the 72 MWe Shippingport Atomic Power Station and an 1175 MWe nuclear plant and the relevancy of the Shippingport decommissioning as a demonstration project for the nuclear industry. For the purpose of this comparison, Portland General Electric Company's 1175 MWe Trojan Nuclear Plant at Rainier, Oregon, has been used as the reference nuclear power plant. 2 refs., 2 figs., 1 tab

  15. Ocean thermal energy conversion (OTEC). Power system development. Preliminary design report, final

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-04

    The preliminary design of the 10 MWe OTEC power module and the 200 kWe test articles is given in detail. System operation and performance; power system cost estimates; 10 MWe heat exchangers; 200 kWe heat exchanger articles; biofouling control;ammonia leak detection, and leak repair; rotating machinery; support subsystem; instrumentation and control; electrical subsystem; installation approach; net energy and resource analysis; and operability, maintainability, and safety are discussed. The conceptual design of the 40 MWe electrical power system includes four or five 10 MWe modules as designed for the 10 MWe pilot plant. (WHK)

  16. Nuclear power: Europe report

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    Last year, 2000, nuclear power plants were available for energy supply, respectively, in 18 countries all over Europe. In eight of the fifteen member countries of the European Union nuclear power plants have been in operation. A total of 218 plants with an aggregate net capacity of 172 259 MWe and an aggregate gross capacity of 181 642 MWe were in operation (31.12.2000; 215 plants, 180 067 MWe (gross), 172 259 MWe (net)). One unit, i.e. Temelin in the Czech Republic went critical for the first time and started test operation after having been connected to the grid. Temelin adds about 981 MWe (gross) and 912 MWe (net) to the electricity production capacity. Three units, Hinkley Point A1 and A2 in United Kingdom, and Chernobyl 3 in the Ukraine have been shut down during the year 2000. This means a loss of 1534 MWe gross capacity and 1420 MWe net capacity. Last year, 12 plants (31.12.2000: 11 plants) were under construction in Romania, Russia, Slovakia, the Czech Republic and the Ukraine, that is only in east european countries. In eight countries of the European Union 146 nuclear power plants have been operated with an aggregate gross capacity of 129 188 MWe and an aggregate net capacity of 123 061 MWe (31.12.2000: 144 plants, 128 613 MWe (gross), 122 627 MWe (net)). Net electricity production in 2000 in the EU amounts to approx. 818.8 TWh, which means a share of 35 per cent of the total production in the whole EU. Shares of nuclear power differ widely among the operator countries. The reach 76 per cent in France, 74 per cent in Lithuania, 57 per cent in Belgium and 47 per cent in the Ukraine. Nuclear power also provides an noticeable share in the electricity supply of countries, which operate no own nuclear power plants, e. g. Italy, Portugal and Austria. (orig.) [de

  17. Analysis of MultiWord Expression Translation Errors in Statistical Machine Translation

    DEFF Research Database (Denmark)

    Klyueva, Natalia; Liyanapathirana, Jeevanthi

    2015-01-01

    In this paper, we analyse the usage of multiword expressions (MWE) in Statistical Machine Translation (SMT). We exploit the Moses SMT toolkit to train models for French-English and Czech-Russian language pairs. For each language pair, two models were built: a baseline model without additional MWE...... data and the model enhanced with information on MWE. For the French-English pair, we tried three methods of introducing the MWE data. For Czech-Russian pair, we used just one method – adding automatically extracted data as a parallel corpus....

  18. An open-label, flexible-dose study of paliperidone extended-release in Chinese patients with first-onset psychosis

    Directory of Open Access Journals (Sweden)

    Si TM

    2015-01-01

    Full Text Available TianMei Si,1 QingRong Tan,2 KeRang Zhang,3 Yang Wang,4 Qing Rui4 1Peking University Institute of Mental health, Key Laboratory of Mental Health, Ministry of Health (Peking University, Beijing, 2Fourth Military Medical University, First Hospital, Xi’an, 3Shanxi Medical University, First Hospital, Shanxi, 4Janssen Research and Development, Beijing, People’s Republic of China Background: Antipsychotic medications facilitate the improvement of psychotic symptoms in patients with first-episode psychosis. Paliperidone extended-release (pali-ER, an atypical anti­psychotic, was assessed for efficacy and safety in Chinese patients with first-episode psychosis. Methods: In this 8-week, open-label, single-arm, multicenter study, patients with first-episode psychosis (Diagnostic and Statistical Manual of Mental Disorders, Fourth Edition criteria and a Positive and Negative Syndrome Scale (PANSS total score ≥70 were treated with flexible-dose pali-ER tablets (3–12 mg/day. The primary efficacy endpoint was the percentage of patients with an increase of ≥8 points in Personal and Social Performance (PSP score from baseline to day 56 (8 weeks. Secondary endpoints included reduction in PANSS total score, improvement in Clinical Global Impression-Severity score, PSP score, Subjective Well-being under Neuroleptics Scale score, and relationship between duration of untreated psychosis and PANSS or PSP. Incidences of treatment-emergent adverse events were used to evaluate safety.Results: Overall, 283 of 294 patients (96% achieved a ≥8-point increase in PSP (primary endpoint, analysis set. For the secondary efficacy endpoints, 284/306 patients (93% had a ≥30% reduction in PANSS total score; 266/306 patients (87% achieved a ≤3 Clinical Global Impression-Severity scale score, and 218/294 patients (74% had a PSP score ≥71. The Subjective Well-being under Neuroleptics Scale score was improved from a baseline mean of 72.7 to 94.7 at endpoint. There was a

  19. Hazards from radioactive waste in perspective

    International Nuclear Information System (INIS)

    Cohen, B.L.

    1979-01-01

    This paper compares the hazards from wastes from a 1000-MW(e) nuclear power plant to these from wastes from a 1000-MW(e) coal fueled power plant. The latter hazard is much greater than the former. The toxicity and carcinogenity of the chemicals prodcued in coal burning is emphasized. Comparisions are also made with other toxic chemicals and with natural radioactivity

  20. Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  1. Operator training and requalification at GPU Nuclear

    International Nuclear Information System (INIS)

    Long, R.L.; Barrett, R.J.; Newton, S.L.

    1982-01-01

    The operator training and requalification programs at GPU Nuclear's Oyster Creek (650 MWe BWR) and Three Mile Island-1 (776 MWe PWR) nuclear plants have undergone significant revisions since the Three Mile Island-2 accident. This paper describes the Training and Education organization, the expanded training facilities, including basic principle trainers and replica simulators, and the present operator training and requalification programs

  2. Landfill lights Liverpool festival

    Energy Technology Data Exchange (ETDEWEB)

    Matan, E

    1986-12-01

    Plants which generate power from garbage landfill gas with outputs up to 10 MWe now run into hundreds around the world. Projects to produce combined-heat-and-power from such resources are relatively few. At Liverpool, UK, a 1 MWe CHP plant has been operating successfully at the site of a major international garden festival.

  3. Present status and future development of Qinshan nuclear power project

    International Nuclear Information System (INIS)

    Ouyang Yu

    1987-01-01

    Qinshan 300 MWe Nuclear power Project is the first domestically designed and constructed nuclear power plant in China. Here is a brief description of its progress in design work, equipment manufacture and site construction since the first structural concrete in March 1985. In Qinshan area four units of 600 MWe each are planned to be built with collaboration of proper foreign partners. (author)

  4. Expert system for assisting the diagnostic and localisation of breakdowns on the fuel elements loading machine

    International Nuclear Information System (INIS)

    Merlin, J.; Pradal, B.

    1990-01-01

    An expert system is developed in order to minimize the time lost through breakdowns of the fuel loading device. The expert system developed by FRAMATOME uses MAINTEX software. The expert systems MACHA and SEDMAC were designed respectively for use on 1300 MWe and 900 MWe loading machines [fr

  5. Cook's Carteaux: Trends in nuclear training

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The following Nuclear News interview, conducted by associate editor Gregg M. Taylor, is with Paul F. Carteaux, training superintendent at Indiana/Michigan Power Company's Cook nuclear power plant. The site has two Westinghouse pressurized water reactors. Cook-1, rated 1020-MWe (net), started commercial operation in August 1975, and the 1060-MWe Cook-2 began operation in July 1978

  6. Data base for a CANDU-PHW operating on a once-through natural uranium cycle

    International Nuclear Information System (INIS)

    1979-07-01

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  7. Central receiver solar thermal power system, phase 1. Progress report for period ending December 31, 1975

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-04-01

    The program objective is the preliminary design of a 10 MWe pilot solar power plant supported by major subsystem experiments. Progress is reported on the following task elements: 10 MWe pilot plant; collector subsystem design and analysis; receiver subsystem requirements; receiver subsystem design; thermal storage subsystem; electrical power generation subsystem; and pilot plant architectural engineering and support. (WDM)

  8. 40 CFR 52.34 - Action on petitions submitted under section 126 relating to emissions of nitrogen oxides.

    Science.gov (United States)

    2010-07-01

    ... group that addressed the problem of ground-level ozone and the long-range transport of air pollution... that had a nameplate capacity greater than 25 MWe and produced electricity for sale under a firm... capacity greater than 25 MWe and produced electricity for sale under a firm contract to the electric grid...

  9. Nuclear Power Prospects

    International Nuclear Information System (INIS)

    Cintra do Prado, L.

    1966-01-01

    The present trend is to construct larger plants: the average power of the plants under construction at present, including prototypes, is 300 MW(e), i.e. three times higher than in the case of plants already in operation. Examples of new large-scale plants ares (a) Wylfa, Anglesey, United Kingdom - scheduled power of 1180 MW(e) (800 MW to be installed by 1967), to be completed in 1968; (b) ''Dungeness B'', United Kingdom - scheduled power of 1200 MW(e); (c) second unit for United States Dresden power plant - scheduled power of 715 MW(e) minimum to almost 800 MW(e). Nuclear plants on the whole serve the same purpose as conventional thermal plants

  10. Nuclear power plants in the world, as of December 31, 1983. 2. ed.

    International Nuclear Information System (INIS)

    1984-01-01

    A List of Nuclear Power Plants all over the world is made every year by JAIF, based on an annual survey on reactors in operation, under construction, on oder, and planned throughout the world. The English version of the List is published now for the second time. The present survey was to find the present status of the world's nuclear power plants as of the end December 1983 as well as changes or new developments during 1983 in the countries listed. The results of the survey are 302 reactors in operation for 198,508.6 MWe, 210 reactors under construction for 205,852 MWe, 13 reactors on order for 10,038 MWe and 134 planned reactors for 134,902 MWe, a total of 659 reactors and a total gross nuclear power generating capacity of 549,300.6 MWe. (author)

  11. Understanding the performance of sulfate reducing bacteria based packed bed reactor by growth kinetics study and microbial profiling.

    Science.gov (United States)

    Dev, Subhabrata; Roy, Shantonu; Bhattacharya, Jayanta

    2016-07-15

    A novel marine waste extract (MWE) as alternative nitrogen source was explored for the growth of sulfate reducing bacteria (SRB). Variation of sulfate and nitrogen (MWE) showed that SRB growth follows an uncompetitive inhibition model. The maximum specific growth rates (μmax) of 0.085 and 0.124 h(-1) and inhibition constants (Ki) of 56 and 4.6 g/L were observed under optimized sulfate and MWE concentrations, respectively. The kinetic data shows that MWE improves the microbial growth by 27%. The packed bed bioreactor (PBR) under optimized sulfate and MWE regime showed sulfate removal efficiency of 62-66% and metals removal efficiency of 66-75% on using mine wastewater. The microbial community analysis using DGGE showed dominance of SRB (87-89%). The study indicated the optimum dosing of sulfate and cheap organic nitrogen to promote the growth of SRB over other bacteria. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. Geothermal potential in Mexico; Potencial geotermico de la republica mexicana

    Energy Technology Data Exchange (ETDEWEB)

    Ordaz Mendez, Christian Arturo; Flores Armenta, Magaly; Ramirez Silva, German [Comision Federal de Electricidad, Gerencia de Proyectos Geotermoelectricos, Morelia, Michoacan (Mexico)]. E-mail: christian.ordaz@cfe.gob.mx

    2011-01-15

    Globally, Mexico is the fourth largest generator of geothermal electricity with an installed capacity of 958 MWe. The Gerencia de Proyectos Geotermoelectricos (GPG, Geothermal-electric division of the Federal Commision for Electricity -CFE) is responsible for using geothermal resources. The GPG calculated the country's geothermal potential as part of CFE's strategy to increase power generation through non-conventional sources. The calculation departed from the GPG's national inventory of thermal manifestations, which is composed of 1380 manifestations scattered throughout the country. At each, surface temperatures were measured and subsurface temperatures estimated by geo-thermometers. The calculation of the geothermal potential was based on the classifying these manifestations by geo-thermometric temperature ranges, providing for high, medium and low enthalpy resources. The volumetric method was used to obtain the national geothermal potential. The results indicate that the Potential Reserves of high-enthalpy resources amounts to 5691 MWe; of moderate-enthalpy resources, 881 MWe; and of low-enthalpy resources, 849 MWe -a total of 7422 MWe. Moreover, the Probable Reserves for high-enthalpy resources amounts to 1643 MWe; of moderate-enthalpy resources, 220 MWe; and of low-enthalpy resources, 212 MWe -a total of 2077 MWe. Finally the Proved Reserves were considered, defined as the additional capacity able to be installed in each known geothermal field, for a total of 186 MWe. All the information was processed and integrated using the Geographic Information System (GIS) ArcGis 9.2 (copyright), resulting in the CFE's intranet publication of the Geothermal Potential Map of Mexico. [Spanish] A nivel mundial, Mexico ocupa el cuarto lugar como generador de electricidad por medio de la energia geotermica con una capacidad instalada de 958 MWe. La Gerencia de Proyectos Geotermoelectricos (GPG) es la responsable del aprovechamiento de estos recursos y como

  13. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  14. Couleur versus noir et blanc

    OpenAIRE

    Boulouch, Nathalie

    2008-01-01

    La couleur n’est entrée que tardivement dans la pratique photographique, avec la production industrielle de la plaque Autochrome Lumière en 1907. Après cette première étape, la commercialisation des procédés à développement chromogène à partir du milieu des années 1930 marquera un nouveau palier, ouvrant l’ère de la photographie couleur moderne. Au-delà du progrès technique généralement retenu, on s’attachera ici au fait que les procédés couleur introduisent une nouvelle catégorie dans le cha...

  15. Tempo 1 méthode de français

    CERN Document Server

    Bérard, Evelyne; Lavenne, Christian

    1996-01-01

    Tempo c'est : une méthode qui s'adresse aux adolescents et adultes, vrais débutants, des objectifs précis, définis par l'observation fine d'échanges authentiques, une progression rigoureuse où les acquis sont constamment repris et élargis, l'acquisition de compétences complètes dans un délai très bref (acquisitions par paliers), des techniques de classe dynamiques : apprentissage par tâches, une mise en scène active de la civilisation, de nombreux exercices, une évaluation régulière, une totale adéquation aux épreuves du DELF.

  16. Commerical electric power cost studies. Capital cost addendum multi-unit coal and nuclear stations

    International Nuclear Information System (INIS)

    1977-09-01

    This report is the culmination of a study performed to develop designs and associated capital cost estimates for multi-unit nuclear and coal commercial electric power stations, and to determine the distribution of these costs among the individual units. This report addresses six different types of 2400 MWe (nominal) multi-unit stations as follows: Two Unit PWR Station-1139 MWe Each, Two Unit BWR Station-1190 MWe Each, Two Unit High Sulfur Coal-Fired Station-1232 MWe Each, Two Unit Low Sulfur Coal-Fired Station-1243 MWe Each, Three Unit High Sulfur Coal-Fired Station-794 MWe Each, Three Unit Low Sulfur Coal-Fired Station-801 MWe Each. Recent capital cost studies performed for ERDA/NRC of single unit nuclear and coal stations are used as the basis for developing the designs and costs of the multi-unit stations. This report includes the major study groundrules, a summary of single and multi-unit stations total base cost estimates, details of cost estimates at the three digit account level and plot plan drawings for each multi-unit station identified

  17. Unit size limitations in smaller power systems

    International Nuclear Information System (INIS)

    McConnach, J.S.

    1975-01-01

    The developing nations have generally found it an economic necessity to accept the minimum commercial size limit of 600 MWe. Smaller reactor sizes tendered as 'one off' specials carry high specific cost penalties which considerably weaken the competitiveness of nuclear versus conventional thermal plants. The revised IAEA market survey for nuclear power in developing countries (1974 edition) which takes account of the recent heavy escalation in oil prices, indicates a reasonable market for smaller size reactors in the range 150 MWe to 400 MWe, but until this market is approached seriously by manufacturers, the commercial availability and economic viability of smaller size reactors remains uncertain. (orig.) [de

  18. Prospects and constraints for nuclear power in developing countries

    International Nuclear Information System (INIS)

    Polliart, A.J.

    1977-01-01

    Despite the interest in nuclear power and the IAEA's active assistance programme, only five developing countries (Argentina, Bulgaria, Czechoslovakia, India, Pakistan) have nuclear plants in operation. The combined net output of these plants is about 2,000 MWe. Twelve other developing countries have nuclear power reactors under construction, ordered or planned for operation by 1985. The net output of those under construction amounts to 17,200 MWe while the ordered or planned reactors will generate an additional 10,300 MWe. (orig./RW) [de

  19. DBE Analysis for KALIMER-600

    International Nuclear Information System (INIS)

    Ha, Kwi Seok; Jeong, Hae Young; Kwon, Young Min; Chang, Won Pyo; Lee, Yong Bum; Kim, Young II

    2009-01-01

    The SFR (Sodium Fast Reactor) which is being developed at KAERI (Korea Atomic Energy Research Institute) is currently divided into three types, such as, Advanced Concept 600 MWe break-even reactor and burner reactor and 1200 MWe break-even reactor. As a part of accidents analysis of the 600 MWe break-even reactor, 5 representative DBE's (Design Bases Events) are analyzed for the safety analysis. The 5 DBE's are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Pipe Break, and SBO (Station Black Out)

  20. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  1. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  2. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Pan Xiren; Zhang Chen.

    1988-01-01

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  3. Performances and reliabilities comparisons by multidimensional statistic studies

    International Nuclear Information System (INIS)

    Coudray, R.

    1993-01-01

    After an overview of the methods utilizables in this type of analyse, we insist on the method used in the experience return for the French PWR 900 MWe chemical and volume control system pump. 7 figs., 3 tabs., 8 refs

  4. Decommissioning and disposal costs in Switzerland

    International Nuclear Information System (INIS)

    Zurkinden, Auguste

    2003-01-01

    Introduction Goal: Secure sufficient financial resources. Question: How much money is needed? Mean: Concrete plans for decommissioning and waste disposal. - It is the task of the operators to elaborate these plans and to evaluate the corresponding costs - Plans and costs are to be reviewed by the authorities Decommissioning Plans and Costs - Comprise decommissioning, dismantling and management (including disposal) of the waste. - New studies 2001 for each Swiss nuclear power plant (KKB 2 x 380 MWe, KKM 370 MWe, KKG 1020 MWe, KKL 1180 MWe). - Studies performed by NIS (D). - Last developments taken into account (Niederaichbach, Gundremmingen, Kahl). Decommissioning: Results and Review Results: Total cost estimates decreasing (billion CHF) 1994 1998 2001 13.7 13.1 11.8 Lower costs for spent fuel conditioning and BE/HAA/LMA repository (Opalinus Clay) Split in 2025: 5.6 bil. CHF paid by NPP 6.2 billion CHF in Fund Review: Concentrates on disposal, ongoing

  5. 2000 MW(t) HTGR-DC-GT Modesto Site dry cooled model 346 concice

    International Nuclear Information System (INIS)

    1979-07-01

    Construction information is presented for a 800 MW(e) HTGR power reactor. The information is itemized for each reactor component or system and incudes quantity, labor hours, labor cost, material cost, and total costs

  6. The world's reactors No. 82: Atucha II

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Detailed information on Atucha II, a 745 MWe pressure-vessel PHWR, is presented in the form of a wall chart. Plant specifications and full colour cutaway drawings of the power station are included. (U.K.)

  7. Iran plans world's fourth biggest nuclear programme

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Power requirements of projected power generation to 1992, and fuel reserves, in Iran are submitted. The current nuclear programme is outlined. 34000 MWe of nuclear power is planned for the end of the century. (U.K.)

  8. Application analysis of Monte Carlo to estimate the capacity of geothermal resources in Lawu Mount

    Energy Technology Data Exchange (ETDEWEB)

    Supriyadi, E-mail: supriyadi-uno@yahoo.co.nz [Physics, Faculty of Mathematics and Natural Sciences, University of Jember, Jl. Kalimantan Kampus Bumi Tegal Boto, Jember 68181 (Indonesia); Srigutomo, Wahyu [Complex system and earth physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha 10, Bandung 40132 (Indonesia); Munandar, Arif [Kelompok Program Penelitian Panas Bumi, PSDG, Badan Geologi, Kementrian ESDM, Jl. Soekarno Hatta No. 444 Bandung 40254 (Indonesia)

    2014-03-24

    Monte Carlo analysis has been applied in calculation of geothermal resource capacity based on volumetric method issued by Standar Nasional Indonesia (SNI). A deterministic formula is converted into a stochastic formula to take into account the nature of uncertainties in input parameters. The method yields a range of potential power probability stored beneath Lawu Mount geothermal area. For 10,000 iterations, the capacity of geothermal resources is in the range of 139.30-218.24 MWe with the most likely value is 177.77 MWe. The risk of resource capacity above 196.19 MWe is less than 10%. The power density of the prospect area covering 17 km{sup 2} is 9.41 MWe/km{sup 2} with probability 80%.

  9. French experience in using programmable systems for the control and the protection of nuclear reactors

    International Nuclear Information System (INIS)

    Jover, P.

    1986-01-01

    The paper presents the results obtained in the use of two automated systems important to safety in 1300 MWE French nuclear reactors: the numerical integrated protection system (SPIN) and the logical control system (CONTROBLOC)

  10. THTR steam generator licensing experience as seen by the manufacturer

    International Nuclear Information System (INIS)

    Fricker, H.W.

    1981-01-01

    This paper describes the licensing procedures of the manufacture of the 300 MWe THTR steam generator. The following problems are discussed: operating data, design, materials used, manufacture and installation of the generator, and also quality control

  11. Investigations on the inadvertent power increase in a PHWR as ASSET experience

    International Nuclear Information System (INIS)

    Kumar, S.H.

    1996-01-01

    Investigations were carried out using the ASSET methodology to find out the root cause of an incident involving inadvertent increase in reactor power in the Unit 1 of Narora Atomic Power Station (NAPS) in India. NAPS is a twin Unit, 220 MWe PHWR based power station. On December 4, 1992, when NPAS Unit 1 was operating at 130 MWe, the reactor power increased steadily on its own and touched 147 MWe, over a period of 14 minutes. The set (demand) power of the triplicated reactor regulating system had increased on its own and in turn has made the reactor to operated at higher power. The power was brought down to 120 MWe by manual intervention. Since adequate system related data during the incident was not available, laboratory studies were carried out using computer simulations for the various process disturbances which could affect the reactor regulating system, for establishing the causes of the event. 4 figs

  12. Investigations on the inadvertent power increase in a PHWR as ASSET experience

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, S H [Operating Plants Safety Div., Atomic Energy Regulatory Board, Mumbai (India)

    1997-12-31

    Investigations were carried out using the ASSET methodology to find out the root cause of an incident involving inadvertent increase in reactor power in the Unit 1 of Narora Atomic Power Station (NAPS) in India. NAPS is a twin Unit, 220 MWe PHWR based power station. On December 4, 1992, when NPAS Unit 1 was operating at 130 MWe, the reactor power increased steadily on its own and touched 147 MWe, over a period of 14 minutes. The set (demand) power of the triplicated reactor regulating system had increased on its own and in turn has made the reactor to operated at higher power. The power was brought down to 120 MWe by manual intervention. Since adequate system related data during the incident was not available, laboratory studies were carried out using computer simulations for the various process disturbances which could affect the reactor regulating system, for establishing the causes of the event. 4 figs.

  13. Two metals welded joints analysis. Specific problems and solution proposal

    International Nuclear Information System (INIS)

    Bodson, F.; Launay, J.P.; Thomas, A.

    1983-03-01

    This paper summarizes the non destructive quality control of bimetallic welded joints on pipes and metallic structures of PWR type reactors (1300 MWe): radiographic and metrasonic failure detection, standardization and in service control processes [fr

  14. Conceptual design of advanced central receiver power systems sodium-cooled receiver concept. Volume 2, Book 2. Appendices. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    The appendices include: (A) design data sheets and P and I drawing for 100-MWe commercial plant design, for all-sodium storage concept; (B) design data sheets and P and I drawing for 100-MWe commercial plant design, for air-rock bed storage concept; (C) electric power generating water-steam system P and I drawing and equipment list, 100-MWe commercial plant design; (D) design data sheets and P and I drawing for 281-MWe commercial plant design; (E) steam generator system conceptual design; (F) heat losses from solar receiver surface; (G) heat transfer and pressure drop for rock bed thermal storage; (H) a comparison of alternative ways of recovering the hydraulic head from the advanced solar receiver tower; (I) central receiver tower study; (J) a comparison of mechanical and electromagnetic sodium pumps; (K) pipe routing study of sodium downcomer; and (L) sodium-cooled advanced central receiver system simulation model. (WHK)

  15. Waste-heat boiler application for the Vresova combined cycle plant

    Energy Technology Data Exchange (ETDEWEB)

    Vicek, Z. [Energoprojekt Praha, Prague (Czechoslovakia)

    1995-12-01

    This report describes a project proposal and implementation of two combined-cycle units of the Vresova Fuel Complex (PKV) with 2 x 200 MWe and heat supply. Participation of ENERGOPROJECT Praha a.s., in this project.

  16. Capital cost: pressurized water reactor plant. Commerical electric power cost studies

    International Nuclear Information System (INIS)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

  17. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  18. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    International Nuclear Information System (INIS)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description

  19. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  20. Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)

    International Nuclear Information System (INIS)

    Sellos, R.F.

    1987-01-01

    It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt

  1. Repowering options for steam power plants

    International Nuclear Information System (INIS)

    Wen, H.; Gopalarathinam, R.

    1992-01-01

    Repowering an existing steam power plant with a gas turbine offers an attractive alternative to a new plant or life extension, especially for unit sizes smaller than 300 MWe. Gas turbine repowering improves thermal efficiency and substantially increases the plant output. Based on recent repowering studies and projects, this paper examines gas turbine repowering options for 100 MWe, 200 MWe and 300 MWe units originally designed for coal firing and currently firing either coal or natural gas. Also discussed is the option for a phased future conversion of the repowered unit to fire coal-derived gas, should there be a fluctuation in the price or availability of natural gas. A modular coal gasification plant designed to shorten the conversion time is presented. Repowering options, performance, costs, and availability impacts are discussed for selected cases

  2. Steam generator in the SNR-project

    International Nuclear Information System (INIS)

    van Westenbrugge, J.K.

    1979-01-01

    The design philosophy of steam generators for 1300 MWe LMFBR's is presented. The basis for this philosophy is the present experience with the licensing of the SNR-300. This experience is reported. The approach for the steam generators for the 1300 MWe LMFBR is elaborated on, both for accident prevention and damage limitation, for the component itself as well as for the system design. Both Design Base Accident and Hypothetical Accidents are discussed. 8 refs

  3. Management of radioactive waste nuclear power plants

    International Nuclear Information System (INIS)

    Dlouhy, Z.; Marek, J.

    1976-01-01

    The authors give a survey of the sources, types and amounts of radioactive waste in LWR nuclear power stations (1,300 MWe). The amount of solid waste produced by a Novovorenezh-type PWR reactor (2 x 400 resp. 1 x 1,000 MWe) is given in a table. Treatment, solidification and final storage of radioactive waste are shortly discussed with special reference to the problems of final storage in the CSR. (HR) [de

  4. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE

  5. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

  6. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  7. Unconventional wind machine

    International Nuclear Information System (INIS)

    Sheff, J.R.

    1979-01-01

    It is the purpose of this paper to introduce an unconventional wind machine which has economics comparable with nuclear power and is already available in the public market place. Specifically, up to about 17 MWE could be saved for other uses such as sale in most 1000 MWE plants of any type - nuclear, oil, gas, peat, or wood - which use conventional electrically driven fans in their cooling towers. 10 refs

  8. Shielding design method for LMFBR validation on the Phenix factor

    International Nuclear Information System (INIS)

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  9. Emission from concentrated sources

    International Nuclear Information System (INIS)

    Ernst, G.

    1981-01-01

    Differential equations to describe the rising cloud of gas and its speed are derived for the case of an inversion layer at constant atmospheric temperature and the case of an indifferent layer of the atmosphere. The characteristics of the cloud of gas at the outlet of a natural draught wet cooling tower of a 1,000 MWe nuclear powerstation and a 600 MWe conventional powerstation are given. (DG) [de

  10. Update on status of fluidized-bed combustion technology

    International Nuclear Information System (INIS)

    Stallings, J.; Boyd, T.; Brown, R.

    1992-01-01

    During the 1980s, fluidized-bed combustion technology has become the dominant technology for solid-fuel-fired power generation systems in the United States. Atmospheric fluidized beds as large as 160 MWe in capacity are now in operation, while pressurized systems reaching 80 MWe have started up in the last year. The commercial status, boiler performance, emissions, and future developments for both atmospheric and pressurized fluidized-bed combustion systems are discussed

  11. Assessment of the Potential of Biomass Gasification for Electricity Generation in Bangladesh

    Directory of Open Access Journals (Sweden)

    Barun Kumar Das

    2014-01-01

    Full Text Available Bangladesh is an agriculture based country where more than 65 percent of the people live in rural areas and over 70% of total primary energy consumption is covered by biomass, mainly agricultural waste and wood. Only about 6% of the entire population has access to natural gas, primarily in urban areas. Electricity production in Bangladesh largely depends on fossil fuel whose reserve is now under threat and the government is now focusing on the alternating sources to harness electricity to meet the continuous increasing demand. To reduce the dependency on fossil fuels, biomass to electricity could play a vital role in this regard. This paper explores the biomass based power generation potential of Bangladesh through gasification technology—an efficient thermochemical process for distributed power generation. It has been estimated that the total power generation from the agricultural residue is about 1178 MWe. Among them, the generation potential from rice husk, and bagasses is 1010 MWe, and 50 MWe, respectively. On the other hand, wheat straw, jute stalks, maize residues, lentil straw, and coconut shell are also the promising biomass resources for power generation which counted around 118 MWe. The forest residue and municipal solid waste could also contribute to the total power generation 250 MWe and 100 MWe, respectively.

  12. Ocean thermal energy conversion power system development-I. Phase I. Preliminary design report. Volume 1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-18

    The results of a conceptual and preliminary design study of Ocean Thermal Energy Conversion (OTEC) closed loop ammonia power system modules performed by Lockheed Missiles and Space Company, Inc. (LMSC) are presented. This design study is the second of 3 tasks in Phase I of the Power System Development-I Project. The Task 2 objectives were to develop: 1) conceptual designs for a 40 to 50-MW(e) closed cycle ammonia commercial plant size power module whose heat exchangers are immersed in seawater and whose ancillary equipments are in a shirt sleeve environment; preliminary designs for a modular application power system sized at 10-MW(e) whose design, construction and material selection is analogous to the 50 MW(e) module, except that titanium tubes are to be used in the heat exchangers; and 3) preliminary designs for heat exchanger test articles (evaporator and condenser) representative of the 50-MW(e) heat exchangers using aluminum alloy, suitable for seawater service, for testing on OTEC-1. The reference ocean platform was specified by DOE as a surface vessel with the heat exchanger immersed in seawater to a design depth of 0 to 20 ft measured from the top of the heat exchanger. For the 50-MW(e) module, the OTEC 400-MW(e) Plant Ship, defined in the Platform Configuration and Integration study, was used as the reference platform. System design, performance, and cost are presented. (WHK)

  13. Nuclear safety in eastern countries. Background of IPSN's actions

    International Nuclear Information System (INIS)

    1999-01-01

    In this document, IPSN presents its opinion about the safety level that might be reached by the nuclear power plants situated in the former-USSR countries. In these countries 2 types of fission reactors are operating: VVER and RBMK with respectively 46 units and 14 units. 3 generations of VVER-type reactors are coexisting: 440 MWe-230, 440 MWe-213 and 1000 MWe-320. The first generation (440 MWe-230) which involves 11 operating units are the least safe and by no means is it possible to make them reach the western standard of safety. The second generation (440 MWe-213) require technical modifications to near western safety standards. The last generation (1000 MWe-320) has safety levels very similar to PWR's if operating procedures are modified and adapted. RBMK-type reactors have been designed in the years 60-70, they suffer from generic defects due to their design, the poor quality of materials and their low reliability. IPSN fears that any incident in such reactors might turn into a major accident. In order to improve nuclear safety in eastern countries, the European Union has launched an international cooperation, the programmes PHARE and TACIS are presented. (A.C.)

  14. Results for Singapore [Market Survey for Nuclear Power in Developing Countries

    International Nuclear Information System (INIS)

    1974-01-01

    A re-analysis of market for nuclear power in Singapore during 1981-1990 was carried out based on the so called ''high'' load forecast of the Market Survey, under the revised economic and technical ground rules, and using the full dynamic programming (DP) optimization option of WASP. In order to reduce computer time requirements the horizon was set at the end of 1995, rather than 2000. During the whole study period, 1977-1995, the DP was allowed free choice of 250 MWe oil-fired plants and of blocks of four 50 MWe gas turbines. During 1981-1995 250 MWe nuclear plants were also allowed to be chosen. During 1985-1995 400 Me nuclear plants were added to the list of candidates, and for 1990-1995 600 MWe nuclear plants were included also. The 250 MWe plant size was chosen because it was in the Public Utility Board expansion plan as the proposed size of units for Senoko Station stage 2 expansion. Gas turbines were considered as expansion candidates in blocks of four, rather than individually, to reduce the number of expansion configuration to be simulated in order to reduce computer time requirements. The 400 and 600 MWe unit sizes were the largest standard size acceptable under the frequency stability criterion in the first year they were permitted as a choice

  15. Fédéralisme et droits des LGBT aux États-Unis et au Canada : analyse comparatives des politiques

    Directory of Open Access Journals (Sweden)

    Miriam Smith

    2014-12-01

    Full Text Available L’article examine le militantisme LGBT et les résultats des politiques en matière de droit pénal et du mariage entre personnes de même sexe de 1969 à aujourd’hui, dans le cadre d’une analyse comparée du fédéralisme aux États-Unis et au Canada. Nous soutiendrons que le fédéralisme a grandement façonné les politiques publiques relatives aux droits des LGBT. Dans le régime américain, le potentiel d’expansion du fédéralisme a favorisé les changements graduels de politiques dans les États et, parallèlement, les multiples points de véto créés par le système de séparation des pouvoirs conjugués aux compétences conférés aux États dans d’importants domaines de politiques ont entravé ces changements de politiques. Dans le régime canadien, la centralisation des mécanismes de protection des droits de la personne dans la Charte de même que la dynamique descendante de l’exercice du pouvoir dans le régime parlementaire de type Westminster ont facilité les changements de politiques. Le fédéralisme est souvent perçu comme un obstacle à l’adoption de politiques progressistes. La répartition des pouvoirs entrave l’élaboration de politiques cohérentes et prête le flanc à la prolifération de vétos dont peuvent se prévaloir des groupes influents pour freiner le changement. Les critiques de cette perspective arguent depuis longtemps que, même si le fédéralisme multiplie les points de véto, il présente aussi de nombreuses possibilités d’innovation. Ils font valoir que les groupes ont plus de poids dans les régimes fédéraux puisqu’ils peuvent exercer leur influence à deux paliers de gouvernement plutôt qu’à un. Ainsi, un groupe qui ne réussit pas à se faire entendre à l’un des paliers pourra tenter sa chance au second et pourra même dresser un palier contre l’autre pour atteindre ses objectifs stratégiques.

  16. Idées nouvelles concernant la variable critique de Leloup. Utilisation de cette variable dans le cas des graisses et dans le cas de l'eau New Ldeas About the Leloup Critical Variable. Use of This Variable for Lubricants and Water

    Directory of Open Access Journals (Sweden)

    Bozet J.

    2006-11-01

    Full Text Available Cet article est essentiellement consacré au fonctionnement. d'un palier lisse en conditions limites, les films édifiés à l'interface étant trop minces pour éviter les interactions de surface. L'auteur montre tout d'abord la relation qui existe entre la variable de Leloup, utilisée par un grand nombre de praticiens, et les caractéristiques du film élasto-hydrodynamique qui se développe aux environs du point critique. Il étudie ensuite le comportement de coussinets graphités lubrifiés à l'eau ; il introduit enfin la lubrification à la graisse de paliers lisses classiques. L'analyse des résultats obtenus au laboratoire permet de conclure à la validité de la variable L pour les deux fluides utilisés. A l'issue de son travail, l'auteur indique que la localisation du point critique relevé expérimentalement dans le cas d'une lubrification à la graisse dépend essentiellement des caractéristiques de l'huile de base, le savon ne jouant qu'un rôle très secondaire. This article mainly deals with the boundary lubrication of journal bearings in which the films at the interfaces are too thin to prevent surface friction. The article first describes the relation existing between the Leloup variable used by a great many technicians and the properties of the elastohydrodynamic film created in the area around the critical point. It then examines the behavior of water-lubricated graphite bearings before going on to the lubrication of conventional journal bearings with grease. An analysis of laboratory results leads to the conclusion that the L variable is valid for both types of fluids. The article concludes by situating the critical point determined experimentally on bearings lubricated with grease. The location of this critical point depends mainly on the properfies of the base oil, in which soap plays only a very secondary role.

  17. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R; Darras, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l'alliage Mg-Zr se comporte nettement mieux que le magnesium pur et surtout que l

  18. Corrosion of magnesium and some magnesium alloys in gas cooled reactors; Corrosion du magnesium et de certains de ses alliages dans les piles refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R.; Darras, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The results of corrosion tests on magnesium and some magnesium alloys (Mg-Zr and Mg-Zr-Zn) in moist air (like G1 reactor) and in CO{sub 2}: (like G2, G3, EDF1 reactors) are reported. The maximum temperature for exposure of magnesium to moist air without any risk of corrosion is 350 deg. C. Indeed, the oxidation rate follows a linear law above 350 deg. C although it reaches a constant level and keeps on very low under 350 deg. C. However, as far as corrosion is concerned this temperature limit can be raised up to 500 deg. C if moist air is very slightly charged with fluorinated compounds. Under pressure of CO{sub 2}, these three materials oxidate much more slowly even if 500 deg. C is reached. The higher is the temperature, the higher is the constant level of the weight increase and the quicker is reached this one. However, Mg-Zr alloy behaves quite better than pure magnesium and especially than Mg-Zr-Zn alloy. (author)Fren. [French] On expose essentiellement les resultats d'etudes sur la corrosion du magnesium et de certains de ses alliages (Mg-Zr et Mg-Zr-Zn) dans l'air humide (cas de la pile G1) et dans le gaz carbonique (cas des piles G2, G3, EDF1, etc...). La temperature limite d'exposition du magnesium dans l'air humide sans risque de corrosion se situe a 350 deg. C; en effet l'oxydation a un caractere lineaire au-dessus de cette temperature, alors qu'elle atteint un palier et reste tres limitee au-dessous de 350 deg. C. Du point de vue de la corrosion, cette temperature limite d'emploi peut cependant etre elevee jusqu'a 500 deg. C si l'on introduit dans l'air humide de tres faibles teneurs de composes fluores. Dans le gaz carbonique sous pression, l'oxydation est beaucoup plus faible, meme jusqu'a 50g. C pour les trois materiaux: l'augmentation de poids atteint un palier d'autant plus eleve et ceci d'autant plus rapidement que la temperature est elle-meme plus elevee. Cependant, l

  19. New nuclear power plants in Europe 1984. Pt 2

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was pubslished in the April 1984 issue, and an overview of the nuclear power plant situation in 26 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. The fifteen nuclear power plants newly commissioned in Europe since last year's atw report was published are covered in a similar way. Moreover, introductory summaries describe the plants in operation in each country and their 1983 electricity generation. A general introduction provides an outlook on developments in Western and Eastern Europe. The total number of nuclear power plants now in operation and under construction in Europe is 300 units with an aggregate gross 219, 320 MWe. Of these, 185 units are located in Western Europe, 115 in Eastern Europe. The units currently in operation of this total are 180 with 103,978 MWe in sixteen European countries; of these, 126 units with 74,869 MWe are run in eleven West European countries, 54 units with an aggregate 29,109 MWe in five East European countries. Of the 120 nuclear generating units at present under construction with an aggregate 115,342 MWe in fifteen European countries, 59 units with 63,442 MWe are located in eight West European, 61 units with 51,900 MWe in seven East European countries. (orig./UA) [de

  20. Planning the nuclear contribution to the Brazilian power program

    International Nuclear Information System (INIS)

    Barbalho, A. Rodrigues; Alves, R. Nazare; Pinto, C. Syllus M.; Souza Santos, T.D. de; Abrao, A.

    1977-01-01

    The thermo-electric power in Brazil accounts for less than 20% of the total generating capacity. Brazil's power is essentially generated hydraulically, which grants the growing development. The electric energy consumption keeps growing, with the annual average rate of 13%/year in the last five years. The present installed capacity is 20 000 MWe and the projections are: 35 000MWe, in 1980; 50 000 MWe in 1985; 75 000 MWe, in 1990 and 150 000 MWe, in 2 000. Most of the hydraulic resources are located in remote areas of the country, very far from the consumption centers. Under the agreement between the Federal Republic of Germany and Brazil (signed in June 1975), besides the nuclear power station, American made, under construction in Agra dos Reis, with a power capacity of 626 MWe, two more units, each one with 1 300 MWe capacity are to be erected at the same site, and planned to be in operation in 1982 and 1983. Several joint German-Brazilian companies will be established for reactor and fuel manufacture, in the country. The Brazilian state holding nuclear company, Empresas Nucleares Brasileiras S/A., NUCLEBRAS, will participate in the formation of all joint companies with at least 51% of capital investments. The Brazilian Government will spend 10 billion dollars (U.S.), during the agreement's duration, to make its industry stronger, to develop its technology and to reduce its dependence on energy imports. Brazil's target: full independence in nuclear technology (including reactor manufacture and complete fuel cycle) in about 15 years [es

  1. Cost and performance of coal-based energy in Brazil

    International Nuclear Information System (INIS)

    Temchin, J.; DeLallo, M.R.

    1998-01-01

    As part of the US Department of Energy's (DOE) efforts to establish the strategic benefits of Clean Coal Technologies (CCT), there is a need to evaluate the specific market potential where coal is a viable option. One such market is Brazil, where significant growth in economic development requires innovative and reliable technologies to support the use of domestic coal. While coal is Brazil's most abundant and economic fossil energy resource, it is presently under utilized in the production of electrical power. This report presents conceptual design for pulverized coal (PC) and circulating fluidized-bed combustion (CFBC) options with resulting capital, operating and financial parameters based on Brazil application conditions. Recent PC and CFBC plant capital costs have dropped with competition in the generation market and have established a competitive position in power generation. Key issues addressed in this study include: Application of market based design approach for FBC and PC, which is competitive within the current domestic, and international power generation markets. Design, fabrication, purchase, and construction methods which reduce capital investment while maintaining equipment quality and plant availability. Impact on coast and performance from application of Brazilian coals, foreign trade and tax policies, construction logistics, and labor requirements. Nominal production values of 200 MWe and 400 MWe were selected for the CFBC power plant and 400 MWe for the PC. The 400 MWe size was chosen to be consistent with the two largest Brazilian PC units. Fluidized bed technology, with limited experience in single units over 200 MW, would consist of two 200 MWe circulating fluidized bed boilers supplying steam to one steam turbine for the 400 MWe capacity. A 200 MWe capacity unit was also developed for CFBC option to support opportunities in re-powering and where specific site or other infrastructure constraints limit production

  2. Regional issue identification and assessment (RIIA). Volume I. An analysis of the TRENDLONG MID-MID Scenario for Federal Region 10

    Energy Technology Data Exchange (ETDEWEB)

    Wilfert, G. L.; Beckwith, M. A.; Cowan, C. E.; Keizur, G. R. [comps.

    1979-07-01

    Environmental, human health and safety, socioeconomic and institutional impacts of future energy development for Federal Region 10, which includes the states of Alaska, Idaho, Oregon, and Washington, are reported. It is concluded that the reduction in electric generating capacity of 568 MWe specified by the scenario for Alaska will not be realized because of institutional constraints and economic impacts. Development of 1000 MWe of geothermal generating capacity in Region 10 called for by the scenario will not be met by 1990. Besides technical feasibility and economic contraints, procedures in Oregon and Washington for securing leases and siting permits have not been fully developed. The location and impacts associated with construction and operation of oil and gas transshipment facilities such as the proposed pipeline to transport natural gas from fields in northern Alaska to the lower 48 states and the pipeline to transport Alaskan oil through Washington State to refineries in the Midwest are likely to be important issues in the Region. The addition of 7,951 MWe to the currently existing hydroelectric generating capacity of 29,990 MWe by 1990 will intensify competition among multiple uses of limited water resources of the Columbia and Snake River systems which drain Idaho, Oregon, and Washington. Irrigation, recreation, transportation, maintenance of wildlife habitats and anadromous fisheries conflict and compete with hydroelectric power generation. Public opposition to further development of nuclear power currently exists and seems to be intensifying in light of recent events. The scenario-specified addition of 4,816 MWe of nuclear generating capacity to the Region's current nuclear capacity of 2,016 MWe may be jeopardized by this opposition; specifically the 1,174 MWe addition to Oregon's nuclear capacity may not be realized.

  3. Realizing vision of Dr. Homi Bhabha - first stage of nuclear power programme

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2009-01-01

    Full text: Dr. Homi Bhabha had a vision to harness nuclear energy for peaceful uses of mankind. Considering typical nuclear resources in the country, Dr Bhabha conceptualized the three stage nuclear power programme and gave a road map for its implementation. The robustness of the vision is such that the programme has not undergone a change in last five decades. The first stage of the three-stage programme, based on natural uranium fuelled Pressurised Heavy Water Reactors (PHWRs) has been precursor to the nuclear power programme in India. This paper describes the developments in the last five decades. The establishment of research laboratories and reactors, training school for the manpower needs, industrial infrastructure and establishment of regulatory framework are briefly described. Setting up of first nuclear power reactor in the country as turnkey project and experience on operation of these reactors in India are discussed. The learning phase consisting of setting up of first PHWR in technical collaboration with Canada and design of 220 MWe PHWRs for MAPS is described. The safety features consistent with development of nuclear power globally were incorporated in Narora design and this became a standardized 220 MWe reactor of which many PHWRs of 220 MWe were set up. The experiences with operation of these small size reactors leading to internationally best operational experience in the year 2002 are discussed. The efforts of plant life extension, in-core maintenance jobs and other renovation and modernization jobs are discussed. The increase in unit size of 540 MWe, of which two reactors have been already set up, is explained in detail. The economies of scale demanded increase in the unit size and design of 700 MWe PHWR has been established and the salient features of this design are also discussed in detail. Eight reactors of 700 MWe each would complete the first stage of about 10,000 MWe PHWR programme and plans for setting up of these reactors are discussed

  4. China: the long wait. [Developments in the nuclear power programme

    Energy Technology Data Exchange (ETDEWEB)

    1990-03-01

    The repercussions of Tiananmen Square coupled with worse than expected balance of payments problems have been made it even more difficult than usual to discern the likely future direction of China's nuclear power programme. Currently there are three power reactors under construction, an indigenously designed 300MWe PWR at Qinsham and the two 900MWe Framatome units at Daya Bay, Guang-dong. Prior to the events of June the most likely way forward, beyond these units under construction, appeared to be development of 600MWe units for Chinese conditions in a joint venture with Western companies. There are strong pressures to ''go indigenous'', and attempt to develop a home grown 600MWe design or, perhaps more likely, go for repeats of the 300 MWe plant - assuming it can be operated successfully, a point on which some factions of the Chinese nuclear industry have their doubts. Nuclear district heating still has strong advocates, as an environmentally sound way of meeting the country's huge projected increases in primary energy consumption over the coming years. Construction of a 5MWt district heating reactor at Tsinghua University, Beijing, started in 1986, entered operation at the end of 1989 following a test phase and is now supplying 3MWt to heating of university buildings. (author).

  5. China: the long wait

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The repercussions of Tiananmen Square coupled with worse than expected balance of payments problems have been made it even more difficult than usual to discern the likely future direction of China's nuclear power programme. Currently there are three power reactors under construction, an indigenously designed 300MWe PWR at Qinsham and the two 900MWe Framatome units at Daya Bay, Guang-dong. Prior to the events of June the most likely way forward, beyond these units under construction, appeared to be development of 600MWe units for Chinese conditions in a joint venture with Western companies. There are strong pressures to ''go indigenous'', and attempt to develop a home grown 600MWe design or, perhaps more likely, go for repeats of the 300 MWe plant - assuming it can be operated successfully, a point on which some factions of the Chinese nuclear industry have their doubts. Nuclear district heating still has strong advocates, as an environmentally sound way of meeting the country's huge projected increases in primary energy consumption over the coming years. Construction of a 5MWt district heating reactor at Tsinghua University, Beijing, started in 1986, entered operation at the end of 1989 following a test phase and is now supplying 3MWt to heating of university buildings. (author)

  6. Market Survey for Nuclear Power in Developing Countries. 1974 Ed. (Preliminary Report)

    International Nuclear Information System (INIS)

    1974-01-01

    In August 1973, the Agency concluded a market survey of nuclear power in fourteen selected developing countries throughout the world. The results of this survey have been reported in individual country reports and in a general report. A summary report on the survey was presented at the seventeenth regular session of the Agency's General Conference. These results indicated that in the fourteen countries surveyed, about 60,000 MWe of nuclear plant capacity might be put into operation during the 1980 to 1989 period. About 94 % of this capacity was represented by units of 600 MWe or larger since under the economic conditions which prevailed in early 1973 nuclear units in the 200 - 400 MWe size range were generally found to be uneconomical compared to oil-fired plants. Following completion of the Market Survey, a very preliminary evaluation of the total potential nuclear market in all developing countries of the world was carried out by Agency staff using the Market Survey results as a basis. This extended study which was completed in late 1973 indicated that the total capacity of nuclear plants which might be installed in the developing world during the 1980 to 1989 period could amount to approximately 160,000 MWe. Of this amount the capacity of small and medium power reactors (200 - 400 MWe) was less than 10,000 MW

  7. On the Effect of Microwave Energy on Lipase-Catalyzed Polycondensation Reactions

    Directory of Open Access Journals (Sweden)

    Alessandro Pellis

    2016-09-01

    Full Text Available Microwave energy (MWe is, nowadays, widely used as a clean synthesis tool to improve several chemical reactions, such as drug molecule synthesis, carbohydrate conversion and biomass pyrolysis. On the other hand, its exploitation in enzymatic reactions has only been fleetingly investigated and, hence, further study of MWe is required to reach a precise understanding of its potential in this field. Starting from the authors’ experience in clean synthesis and biocatalyzed reactions, this study sheds light on the possibility of using MWe for enhancing enzyme-catalyzed polycondensation reactions and pre-polymer formation. Several systems and set ups were investigated involving bulk and organic media (solution phase reactions, different enzymatic preparations and various starting bio-based monomers. Results show that MWe enables the biocatalyzed synthesis of polyesters and pre-polymers in a similar way to that reported using conventional heating with an oil bath, but in a few cases, notably bulk phase polycondensations under intense microwave irradiation, MWe leads to a rapid enzyme deactivation.

  8. MHD generator performance analysis for the Advanced Power Train study

    Science.gov (United States)

    Pian, C. C. P.; Hals, F. A.

    1984-01-01

    Comparative analyses of different MHD power train designs for early commercial MHD power plants were performed for plant sizes of 200, 500, and 1000 MWe. The work was conducted as part of the first phase of a planned three-phase program to formulate an MHD Advanced Power Train development program. This paper presents the results of the MHD generator design and part-load analyses. All of the MHD generator designs were based on burning of coal with oxygen-enriched air preheated to 1200 F. Sensitivities of the MHD generator design performance to variations in power plant size, coal type, oxygen enrichment level, combustor heat loss, channel length, and Mach number were investigated. Basd on these sensitivity analyses, together with the overall plant performance and cost-of-electricity analyses, as well as reliability and maintenance considerations, a recommended MHD generator design was selected for each of the three power plants. The generators for the 200 MWe and 500 MWe power plant sizes are supersonic designs. A subsonic generator design was selected for the 1000 MWe plant. Off-design analyses of part-load operation of the supersonic channel selected for the 200 MWe power plant were also conductd. The results showed that a relatively high overall net plant efficiency can be maintained during part-laod operation with a supersonic generator design.

  9. Nuclear power: 2004 world report - evaluation

    International Nuclear Information System (INIS)

    Anon.

    2005-01-01

    Last year, 2004, 441 nuclear power plants were available for power supply in 31 countries of the world. Nuclear generating capacity attained its highest level so far at an aggregate gross power of 385,854 MWe and an aggregate net power of 366,682 MWe, respectively. Nine different reactor lines are operated in commercial nuclear power plants. Light water reactors (PWR and BWR) again are in the lead with 362 plants. At year's end, 22 nuclear power plants with an aggregate gross power of 18,553 MWe and an aggregate net power, respectively, of 17,591 MWe were under construction in nine countries. Of these, twelve are light water reactors, nine are CANDU-type reactors, and one is a fast breeder reactor. So far, 104 commercial reactors with powers in excess of 5 MWe have been decommissioned in eighteen countries, most of them low-power prototype plants. 228 nuclear power plants of those in operation, i.e. slightly more than half, were commissioned in the 1980es. Nuclear power plant availabilities in terms of capacity and time again reached record levels. Capacity availability was 84.30%, availability in terms of time, 85.60%. The four nuclear power plants in Finland continue to be world champions in this respect with a cumulated average capacity availability of 90.30%. (orig.)

  10. Proposition de Mesures Magnétiques pour le Projet AD (hormis les dipôles de correction)

    CERN Document Server

    Cornuet, D

    1997-01-01

    Pour le projet AD (Antiproton Decelerator), c'est la machine AC (Antiproton Collector) qui va être utilisée. Mais il s'agit non plus de fonctionner à courant constant (correspondant à 3,57 GeV/c) mais de décélérer le faisceau d'antiprotons à courant variable et stabilisation sur différents paliers décroissants (3,57 GeV/c, 2 GeV/c, 300 MeV/c pour le refroidissement) afin d'éjecter à basse énergie (correspondant à 100 MeV/c). Les aimants de la machine AC avaient été shimmés et optimisés pour 3,57 GeV/c (références 1 et 2). Dans ce nouveau contexte de la machine AD, il convient de vérifier si un compromis des shimmings existants ne doit pas être trouvé pour obtenir une homogénéité de champ à toutes énergies. Il faut vérifier aussi que les courants de Foucault induits dans les chambres à vide rectangulaires ne perturbent pas la qualité de champ. La possibilité d'utiliser des sondes NMR comme marqueurs de train B est aussi étudiée.

  11. 4{pi} proportional counter for absolute measurement of {beta}-emitters; Compteur 4{pi} proportionnel destine a la mesure absolue d'emetteurs {beta}

    Energy Technology Data Exchange (ETDEWEB)

    Perolat, J P; Laine, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    The 4{pi} counter is commonly used to measure the activity of {beta} sources, the one we describe does not advance a new conception but the issue of several years of investigations. The apparatus presents good performances about the qualities commonly required in that kind of counter: detection efficiency, plateau and dead time. Moreover technical characteristics give a great flexibility in use, particularly a possibility of adaptation in a {beta},{gamma} coincidence unit. (authors) [French] Le compteur 4{pi} proportionnel est couramment, utilise pour mesurer l'activite de sources {beta}, celui que nous decrivons ne represente pas une conception nouvelle, mais plutot l'aboutissement de plusieurs annees d'etudes. Cette installation presente des performances satisfaisantes en ce qui concerne les qualites habituellement requises dans ce type de compteur: l'efficacite de detection, le palier et le temps mort. En outre, certaines caracteristiques techniques qui lui sont propres lui conferent une grande souplesse d'utilisation, en particulier la possibilite d'adaptation a un ensemble de mesures par coincidences 4{pi} {beta} {gamma}. (auteurs)

  12. Les enjeux éthiques de l’élaboration des politiques publiques favorables à la santé : réflexion et illustration

    Directory of Open Access Journals (Sweden)

    Danielle Blondeau

    2011-06-01

    Full Text Available Au Québec, en vertu de la Loi sur la santé publique (L.R.Q. chapitre S-2.2, 2001, article 54, le ministre de la Santé a un droit de regard sur les mesures prévues par les autres ministères dans leurs lois et règlements qui pourraient avoir un impact significatif sur la santé de la population. Ainsi, les acteurs de santé publique du palier central ont cherché à soutenir la mise en oeuvre de cet article, de manière à favoriser l’adoption de politiques publiques favorables à la santé. Ce texte examine les enjeux éthiques liés à l’élaboration de telles politiques. Après avoir exposé les tensions entre bien individuel et bien commun ainsi que les valeurs de justice sociale, responsabilité et démocratie qui y sont reliées, nous illustrerons les enjeux éthiques à travers l’exemple du cinémomètre photographique comme mesure visant à augmenter la sécurité routière.

  13. Still waiting for the green light on Taiwan's units 7 and 8

    International Nuclear Information System (INIS)

    Lin, E.

    1992-01-01

    Taiwan Power Company (Taipower) is the only utility supplying electricity to Taiwan. In 1991, six nuclear units shared 28% of the total installed capacity (5144MWe out of 18 382MWe), but produced 38% of the total electricity (33 878TWh out of 89 129TWh), with a 7% increase over 1990. The weighted average capacity factor reached a record high of 78.32%. Compared with 1990's weighted average capacity factor of 72.94%, the annual performance in 1991 reveals that Taipower nuclear power plants are in better shape than they were before. The major improvement efforts in 1992 will focus on shortening the duration of outages and enforcing safety culture training. This article also briefly describes existing and projected waste management plants and comments on the project to build two 1000MWe Light Water Reactor plants at Yenliao which are tentatively scheduled for commercial operation in 2000. (Author)

  14. Integrity of pressurized water electronuclear reactor vessels. The case of French reactors

    International Nuclear Information System (INIS)

    2012-01-01

    This document aims at identifying elements related to design, manufacturing and control during operation of reactor vessels of the French electronuclear fleet, and more precisely as far as vessel ferrule is concerned. It briefly describes the typical design and elements of most of French PWR vessels with respect to the reactor type (900 MWe, 1300 MWe, 1450 MWe, EPR). It recalls some measures regarding design (for embrittlement assessment) and manufacturing processes (forging operations for shell fabrication, coatings). It discusses the different manufacturing defects which have been noticed (under the coatings, due to hydrogen, and intergranular loss of cohesion due to re-heating). It more particularly comments defects noticed on a Belgium power station reactor in Doel, defects due to hydrogen and some other defects noticed in the French reactor fleet. It presents the different types of control which are performed on vessel shells during operation

  15. Progress in the developing countries

    International Nuclear Information System (INIS)

    Simnad, M.

    1981-01-01

    Nuclear programmes in selective developing countries are briefly discussed. The oil rich countries of Iraq, Libya and Iran all have reactors on order. Turkey has decided to purchase a PWR from the USSR and Egypt's programme anticipates a capacity of 6600 MWe by 2000. The current projections for India are 6000 MWe by 1990 and 20,000 MWe by 2000. The progress of Pakistan, South Korea and other Asian countries are discussed. The predicted growth in reactors and population in Latin America is considered - 17 reactors presently planned for a population of 340 million and 18-57 possible additions in 2000 for an estimated population of 600 million. The role of the IAEA and experience of some Western countries in technology transfer is discussed with the ambitious Spanish nuclear power programme and the experience of Argentina in purchasing Candu reactors. (author)

  16. The costs of nuclear power in the Netherlands

    International Nuclear Information System (INIS)

    1978-01-01

    A study on the costs of nuclear power generation in the Netherlands is presented. Light water cooled reactors are chosen as nuclear power plants and no difference is made in calculating the costs between a PWR type reactor and a BWR type reactor. The power plants have an output of 1000 MWe. From each part of the whole fuel cycle the costs are determined, taking into account interest, investments, time of construction, labor costs, insurances etc. Also are determined from each part of the fuel cycle the energy costs; the costs per kWh. Finally a comparison is made in costs between a 1000 MWe power plant and a 600 MWe power plant

  17. Portland General Electric Company report on the operating and startup experience with control and instrumentation and electrical systems at the Trojan Nuclear Plant

    International Nuclear Information System (INIS)

    Zimmerman, G.A.

    1977-01-01

    The Trojan Nuclear Plant is an 1178 MWe nuclear plant located on the Columbia River 40 miles northwest of Portland, Oregon. The Nuclear Stream Supply System vendor is Westinghouse with a General Electric turbine generator. The reactor is rated and licensed for 3423 MWt (1178 MWe) and the turbine generator is designed for 3570 MWt(1219 MWe). The startup phase testing of Trojan commenced on November 21, 1975, upon receipt of our NRC Operating License. The startup testing program was completed on May 22, 1976, following 100 hours of full-power operation, at which time a scheduled summer maintenance outage began. Some of the highlights and milestones of the startup testing program are described

  18. Laser fusion hybrid reactor systems study

    International Nuclear Information System (INIS)

    1976-07-01

    The work was performed in three phases. The first phase included a review of the many possible laser-reactor-blanket combinations and resulted in the selection of a ''demonstration size'' 500 MWe plant for further study. A number of fast fission blankets using uranium metal, uranium-molybdenum alloy, and uranium carbide as fuel were investigated. The second phase included design of the reactor vessel and internals, heat transfer system, tritium processing system, and the balance of plant, excluding the laser building and equipment. A fuel management scheme was developed, safety considerations were reviewed, and capital and operating costs were estimated. Costs developed during the second phase were unexpectedly high, and a thorough review indicated considerable unit cost savings could be obtained by scaling the plant to a larger size. Accordingly, a third phase was added to the original scope, encompassing the redesign and scaling of the plant from 500 MWe to 1200 MWe

  19. Staging Rankine Cycles Using Ammonia for OTEC Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Bharathan, D.

    2011-03-01

    Recent focus on renewable power production has renewed interest in looking into ocean thermal energy conversion (OTEC) systems. Early studies in OTEC applicability indicate that the island of Hawaii offers a potential market for a nominal 40-MWe system. a 40-MWe system represents a large leap in the current state of OTEC technology. Lockheed Martin Inc. is currently pursuing a more realistic goal of developing a 10-MWe system under U.S. Navy funding (Lockheed 2009). It is essential that the potential risks associated with the first-of-its-kind plant should be minimized for the project's success. Every means for reducing costs must also be pursued without increasing risks. With this in mind, the potential for increasing return on the investment is assessed both in terms of effective use of the seawater resource and of reducing equipment costs.

  20. Extending the operating lifetime of the nuclear power plants in France

    International Nuclear Information System (INIS)

    Ancelin, C.

    2015-01-01

    In France 58 reactors were deployed between 1977 and 2000, they are from only 3 standardized series: 900 MWe (34 units), 1300 MWe (20 units) and 1500 MWe (4 units). The average age of the reactor fleet is 29 years. This series of slides details the EDF's strategy for extending plant lifetime significantly beyond 40 years. An important point is the management and anticipation of the ageing of equipment through the distinction between replaceable components and non-replaceable components (mainly pressure vessel and containment building) and by listing all the deterioration ways possible. A second important point is a large scale research program to demonstrate the fitness for service of non-replaceable equipment after 40 years. This program focuses on physical modelling, computerized simulations and improving non-destructive techniques. (A.C.)

  1. Reducing costs by reducing size

    International Nuclear Information System (INIS)

    Hayns, M.R.; Shepherd, J.

    1991-01-01

    The present paper discusses briefly the many factors, including capital cost, which have to be taken into account in determining whether a series of power stations based on a small nuclear plant can be competitive with a series based on traditional large unit sizes giving the guaranteed level of supply. The 320 MWe UK/US Safe Integral Reactor is described as a good example of how the factors discussed can be beneficially incorporated into a design using proven technology. Finally it goes on to illustrate how the overall costs of a generating system can indeed by reduced by use of the 320 MWe Safe Integral Reactor rather than conventional units of around 1200 MWe. (author). 9 figs

  2. IAEA’s Perspectives on Global Nuclear Power – Opportunities and Challenges

    International Nuclear Information System (INIS)

    Park, J.K.

    2014-01-01

    Status of global nuclear power: 437 reactors in operation (374.5 GWe); 2 reactors in long-term shutdown; 149 reactors in permanent shutdown; 70 reactors under construction. [As of Sep. 2014] Latest connections to the grid: - Ningde-2, 1000 MW(e), PWR, China; - Atucha-2, 692 MW(e), PHWR, Argentina; - Fuqing-1, 1000 MW(e), PWR, China). [Website: http://www.iaea.org/pris/]. IAEA projections of nuclear power: • Sep. 2014: 374.5 GWe; • 2030 - low 400.6 GWe: 7.0% increase; - high 699.2 GWe: 86.7% increase; • 2050 - low 412.9 GWe: 10.3% increase; - high 1091.7 GWe: 191.5% increase

  3. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  4. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, A. dos; Nascimento, J.A. do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  5. Evolution of the roof insulation conception in French LMFBR

    International Nuclear Information System (INIS)

    Frachet, S.; Pradel, P.

    1986-01-01

    France has built two power producing fast breeders: the 250 MWe PHENIX and the 1200 MWe SUPER-PHENIX-1 (with European participation) and is presently working on the preliminary design of a third 1500 MWe reactor SUPER-PHENIX-2. The upper closures of these pool type reactors are essential structural elements since they fulfill several vital functions. The following points are briefly discussed in this paper: The different facilities which have contributed to validate the SPX1 and SPX2 upper-shuttings: Gulliver, Coca, Germinal; the numerical methods which were perfected to validate the calculation of the penetrations; a heat balance which is used to estimate the temperature of the cover gas; the solution adopted to avoid the aerosols deposits

  6. Status of the R and D activities on fast reactors and accelerators driven system in Brazil

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  7. Comparative of fuel cycle cost for light water nuclear power plants; Uporedna analiza cene gorivnog ciklusa lakovodne nuklearne elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Dimitrijevic, Z [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  8. Preliminary core design of IRIS-50

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Franceschini, Fausto

    2009-01-01

    IRIS-50 is a small, 50 MWe, advanced PWR with integral primary system. It evolved employing the same design principles as the well known medium size (335 MWe) IRIS. These principles include the 'safety-by-design' philosophy, simple and robust design, and deployment flexibility. The 50 MWe design addresses the needs of specific applications (e.g., power generation in small regional grids, water desalination and biodiesel production at remote locations, autonomous power source for special applications, etc.). Such applications may favor or even require longer refueling cycles, or may have some other specific requirements. Impact of these requirements on the core design and refueling strategy is discussed in the paper. Trade-off between the cycle length and other relevant parameters is addressed. A preliminary core design is presented, together with the core main reactor physics performance parameters. (author)

  9. Working group 4a: Regional aspects. Nuclear power plants siting in the dutch speaking part of the country

    International Nuclear Information System (INIS)

    Willems, M.; Medart, R.; Vanneste, O.

    1976-01-01

    The problems due to nuclear plant siting in the northern region of Belgium are reviewed with an emphasis on economical, environmental and esthetical aspects. Three types of sitings were analysed: inland, coastal and off-shore. For the in-land siting, Doel, where already two units are in operation (780 MWe) and a third in construction (900 MWe), is supposed to be able to receive a fourth unit of 1000 MWe. The coastal siting is practically impossible for two reasons: the lack of cooling water when a coastal inland region of 5 km is considered and the strong density of tourists on the 66 km coast. For artificial island siting the different aspects are considered: type of soil, marine environment, construction factors, security, construction time, costs, etc. A comparative study for 9 off-shore sites is presented. (A.F.)

  10. Core design and fuel management studies

    International Nuclear Information System (INIS)

    Min, Byung Joo; Chan, P.

    1997-06-01

    The design target for the CANDU 9 requires a 20% increase in electrical power output from an existing 480-channel CANDU core. Assuming a net electrical output of 861 MW(e) for a natural uranium fuelled Bruce-B/Darlington reactor in a warm water site, the net electrical output of the reference CANDU 9 reactor would be 1033 MW(e). This report documents the result of the physics studies for the design of the CANDU 9 480/SEU core. The results of the core design and fuel management studies of the CANDU 9 480/SEU reactor indicated that up to 1033 MW(e) output can be achieved in a 480-channel CANDU core by using SEU core can easily be maintained indefinitely using an automated refuelling program. Fuel performance evaluation based on the data of the 500 FPDs refuelling simulation concluded that SEU fuel failure is not expected. (author). 2 tabs., 38 figs., 5 refs

  11. Comparative of fuel cycle cost for light water nuclear power plants

    International Nuclear Information System (INIS)

    Kocic, A.; Dimitrijevic, Z.

    1978-01-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  12. Evaluation of the commercial FBR introduction date

    International Nuclear Information System (INIS)

    White, M.K.; Merrill, E.T.

    1981-09-01

    This report examines one criterion for introducing a commercial FBR: economic competitiveness with a Light Water Reactor (LWR). For this analysis, the commercial FBR is assumed to be the fifth-of-a kind replicate which represents an economically mature plant. This FBR is deemed economically competitive when its life-cycle energy cost is less than or equal to that of an LWR. Results of this analysis are used in a comparative analysis of alternative FBR development stategies. The strategies evaluated in these studies assume both 1000- and 1457-MWe FBRs. Since the capital costs per kilowatt, and therefore the energy costs, for these two FBR sizes are different, they will become economically competitive at different times. The probability density function for the 1457-MW(e) FBR has an expected value date or weighted average date of 2030, compared with 2033 for the probability density function for the 1000-MW(e) FBR

  13. Nuclear power world report 2013

    International Nuclear Information System (INIS)

    Anon.

    2014-01-01

    At the end of 2013, 435 nuclear power plants were available for energy supply in 31 countries of the world. This means that the number decreased by 2 units compared to the previous year's number on 31 December 2012. The aggregate gross power of the plants amounted to approx. 398,861 MWe, the aggregate net power, to 378,070 MWe (gross: 392,793 MWe, net: 372,572 MWe, new data base as of 2013: nameplate capacities). Four units were commissioned in 2014; three units in China and one in India. Eight units were shut down permanently in 2013; 2 units in Japan, and four units in the USA. Two units in Canada were declared permanently shut-down after a long-term shutdown. 70 nuclear generating units - 2 more than at the end of 2012 - were under construction in late 2013 in 15 countries with an aggregate gross power of approx. 73,814 MWe and net power of approx. 69,279 MWe. Six new projects have been started in 2013 in four countries (Belarus, China, the Republic of Korea, and the United Arab Emirates). Worldwide, some 125 new nuclear power plants are in the concrete project design, planning, and licensing phases; in some of these cases license applications have been submitted or contracts have already been signed. Some 100 further projects are planned. Net electricity generation in nuclear power plants worldwide in 2013 achieved a level of approx. 2,364.15 billion (109) kWh (2012: approx. 2,350.80 billion kWh). Since the first generation of electricity in a nuclear power plant in the EBR-I fast breeder (USA) on December 20, 1951, cumulated net production has reached approx. 70,310 billion kWh, and operating experience has grown to some 15,400 reactor years. (orig.)

  14. Investigation of small and modular-sized fast reactor

    International Nuclear Information System (INIS)

    Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

    2000-06-01

    In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

  15. Nuclear power world report 2013

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2014-07-15

    At the end of 2013, 435 nuclear power plants were available for energy supply in 31 countries of the world. This means that the number decreased by 2 units compared to the previous year's number on 31 December 2012. The aggregate gross power of the plants amounted to approx. 398,861 MWe, the aggregate net power, to 378,070 MWe (gross: 392,793 MWe, net: 372,572 MWe, new data base as of 2013: nameplate capacities). Four units were commissioned in 2014; three units in China and one in India. Eight units were shut down permanently in 2013; 2 units in Japan, and four units in the USA. Two units in Canada were declared permanently shut-down after a long-term shutdown. 70 nuclear generating units - 2 more than at the end of 2012 - were under construction in late 2013 in 15 countries with an aggregate gross power of approx. 73,814 MWe and net power of approx. 69,279 MWe. Six new projects have been started in 2013 in four countries (Belarus, China, the Republic of Korea, and the United Arab Emirates). Worldwide, some 125 new nuclear power plants are in the concrete project design, planning, and licensing phases; in some of these cases license applications have been submitted or contracts have already been signed. Some 100 further projects are planned. Net electricity generation in nuclear power plants worldwide in 2013 achieved a level of approx. 2,364.15 billion (109) kWh (2012: approx. 2,350.80 billion kWh). Since the first generation of electricity in a nuclear power plant in the EBR-I fast breeder (USA) on December 20, 1951, cumulated net production has reached approx. 70,310 billion kWh, and operating experience has grown to some 15,400 reactor years. (orig.)

  16. The present state of the HTR concept based on experience gained from AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, W.

    1989-01-01

    During the past ten years the development of a specific HTR concept has made remarkable progress. This has been mainly characterized by making use of the safety characteristics typical of the High-Temperature Reactor (HTR). In the design, construction and operation of High-Temperature Reactors - especially AVR (15 MWe plant in Juelich, FRG) and THTR (300 MWe plant in Hamm-Uentrop, FRG) - comprehensive experience has been gained in the field of operational availability and safety, accident topology and plant risk of HTRs in recent years. This experience is relevant for the entire HTR line independent of specific projects. (author). 3 refs, 5 figs, 1 tab

  17. The control of emissions from nuclear power reactors in Canada

    International Nuclear Information System (INIS)

    Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

    1988-01-01

    Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

  18. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  19. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

  20. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  1. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    International Nuclear Information System (INIS)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices

  2. Parametric design study of tandem mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.

    1977-01-01

    The parametric design study of the tandem mirror reactor (TMR) is described. The results of this study illustrate the variation of reactor characteristics with changes in the independent design parameters, reveal the set of design parameters which minimizes the cost of the reactor, and show the sensitivity of the optimized design to physics and technological uncertainties. The total direct capital cost of an optimized 1000 MWe TMR is estimated to be $1300/kWe. The direct capital cost of a 2000 MWe plant is less than $1000/kWe

  3. The United States of America Country Update

    Energy Technology Data Exchange (ETDEWEB)

    Lund, John W. (1); Bloomquist, R. Gordon (2); Boyd, Tonya L. (1); Renner, Joel (3); (1) Geo-Heat Center, Oregon Institute of Technology, Klamath Falls, OR; (2) Washington State University Energy Program, Olympia, WA; (3) Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID

    0001-01-01

    Geothermal energy is used for electric power generation and direct utilization in the United States. The present installed capacity (gross) for electric power generation is 2,534 MWe with about 2,000 MWe net delivering power to the grid producing approximately 17,840 GWh per year for a 80.4% gross capacity factor. Geothermal electric power plants are located in California, Nevada, Utah and Hawaii. The two largest concentrations of plants are at The Geysers in northern California and the Imperial Valley in southern California. The latest development at The Geysers, starting in 1998, is injecting recycled wastewater from two communities into the reservoir, which presently has recovered about 100 MWe of power generation. The second pipeline from the Santa Rosa area has just come on line. The direct utilization of geothermal energy includes the heating of pools and spas, greenhouses and aquaculture facilities, space heating and district heating, snow melting, agricultural drying, industrial applications and groundsource heat pumps. The installed capacity is 7,817 MWt and the annual energy use is about 31,200 TJ or 8,680 GWh. The largest application is ground-source (geothermal) heat pumps (69% of the energy use), and the next largest direct-uses are in space heating and agricultural drying. Direct utilization (without heat pumps) is increasing at about 2.6% per year; whereas electric power plant development is almost static, with only about 70 MWe added since 2000 (there were errors in the WGC2000 tabulation). A new 185-MWe plant being proposed for the Imperial Valley and about 100 MWe for Glass Mountain in northern California could be online by 2007-2008. Several new plants are proposed for Nevada totaling about 100 MWe and projects have been proposed in Idaho, New Mexico, Oregon and Utah. The total planned in the next 10 years is 632 MWe. The energy savings from electric power generation, direct-uses and ground-source heat pumps amounts to almost nine million tonnes

  4. Solar thermal repowering systems integration. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dubberly, L. J.; Gormely, J. E.; McKenzie, A. W.

    1979-08-01

    This report is a solar repowering integration analysis which defines the balance-of-plant characteristics and costs associated with the solar thermal repowering of existing gas/oil-fired electric generating plants. Solar repowering interface requirements for water/steam and salt or sodium-cooled central receivers are defined for unit sizes ranging from 50 MWe non-reheat to 350 MWe reheat. Finally balance-of-plant cost estimates are presented for each of six combinations of plant type, receiver type and percent solar repowering.

  5. The United States of America Country Update

    Energy Technology Data Exchange (ETDEWEB)

    Lund, John W [1; Bloomquist, R Gordon [2; Boyd, Tonya L [1; Renner, Joel [3; (1) Geo-Heat Center, Oregon Institute of Technology, Klamath Falls, OR; (2) Washington State University Energy Program, Olympia, WA; (3) Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID

    0000-12-30

    Geothermal energy is used for electric power generation and direct utilization in the United States. The present installed capacity (gross) for electric power generation is 2,534 MWe with about 2,000 MWe net delivering power to the grid producing approximately 17,840 GWh per year for a 80.4% gross capacity factor. Geothermal electric power plants are located in California, Nevada, Utah and Hawaii. The two largest concentrations of plants are at The Geysers in northern California and the Imperial Valley in southern California. The latest development at The Geysers, starting in 1998, is injecting recycled wastewater from two communities into the reservoir, which presently has recovered about 100 MWe of power generation. The second pipeline from the Santa Rosa area has just come on line. The direct utilization of geothermal energy includes the heating of pools and spas, greenhouses and aquaculture facilities, space heating and district heating, snow melting, agricultural drying, industrial applications and groundsource heat pumps. The installed capacity is 7,817 MWt and the annual energy use is about 31,200 TJ or 8,680 GWh. The largest application is ground-source (geothermal) heat pumps (69% of the energy use), and the next largest direct-uses are in space heating and agricultural drying. Direct utilization (without heat pumps) is increasing at about 2.6% per year; whereas electric power plant development is almost static, with only about 70 MWe added since 2000 (there were errors in the WGC2000 tabulation). A new 185-MWe plant being proposed for the Imperial Valley and about 100 MWe for Glass Mountain in northern California could be online by 2007-2008. Several new plants are proposed for Nevada totaling about 100 MWe and projects have been proposed in Idaho, New Mexico, Oregon and Utah. The total planned in the next 10 years is 632 MWe. The energy savings from electric power generation, direct-uses and ground-source heat pumps amounts to almost nine million tonnes

  6. Distribution of arrival times of muons with energy greater than 10 GeV

    International Nuclear Information System (INIS)

    Badino, G.; Bianco, P.; Dardo, M.; Fulgione, W.; Galeotti, P.; Periale, L.; Saavedra, O.

    1982-01-01

    Recent data on the arrival time distribution of EAS of primary energy >=10 14 eV, and of high energy muons detected at great depth (5000 mwe), seem to indicate an excess of short time intervals. We are using an apparatus, installed at 40 mwe underground, and a surface shower array to investigate the distributions of a) the time intervals between muon groups and b) the arrival times of muons with respect to the front of air showers. Preliminary results of this search are presented

  7. Ontario Hydro decontamination experience

    Energy Technology Data Exchange (ETDEWEB)

    Lacy, C S; Patterson, R W; Upton, M S [Chemistry and Metallurgy Department, Central Production Services, Ontario Hydro, ON (Canada)

    1991-04-01

    Ontario Hydro currently operates 18 nuclear electric generating units of the CANDU design with a net capacity of 12,402 MW(e). An additional 1,762 MW(e) is under construction. The operation of these facilities has underlined the need to have decontamination capability both to reduce radiation fields, as well as to control and reduce contamination during component maintenance. This paper presents Ontario Hydro decontamination experience in two key areas - full heat transport decontamination to reduce system radiation fields, and component decontamination to reduce loose contamination particularly as practised in maintenance and decontamination centres. (author)

  8. The United States of America country update

    Energy Technology Data Exchange (ETDEWEB)

    Lund, John W.; Bloomquist, R. Gordon; Boyd, Tonya L.; Renner, Joel

    2005-01-01

    Geothermal energy is used for electric power generation and direct utilization in the United States. The present installed capacity (gross) for electric power generation is 2,534 MWe with about 2,000 MWe net delivering power to the grid producing approximately 17,840 GWh per year for a 80.4% gross capacity factor. Geothermal electric power plants are located in California, Nevada, Utah and Hawaii. The two largest concentrations of plants are at The Geysers in northern California and the Imperial Valley in southern California. The latest development at The Geysers, starting in 1998, is injecting recycled wastewater from two communities into the reservoir, which presently has recovered about 100 MWe of power generation. The second pipeline from the Santa Rosa area has just come on line. The direct utilization of geothermal energy includes the heating of pools and spas, greenhouses and aquaculture facilities, space heating and district heating, snow melting, agricultural drying, industrial applications and groundsource heat pumps. The installed capacity is 7,817 MWt and the annual energy use is about 31,200 TJ or 8,680 GWh. The largest application is ground-source (geothermal) heat pumps (69% of the energy use), and the next largest direct-uses are in space heating and agricultural drying. Direct utilization (without heat pumps) is increasing at about 2.6% per year; whereas electric power plant development is almost static, with only about 70 MWe added since 2000 (there were errors in the WGC2000 tabulation). A new 185-MWe plant being proposed for the Imperial Valley and about 100 MWe for Glass Mountain in northern California could be online by 2007-2008. Several new plants are proposed for Nevada totaling about 100 MWe and projects have been proposed in Idaho, New Mexico, Oregon and Utah. The total planned in the next 10 years is 632 MWe. The energy savings from electric power generation, direct-uses and ground-source heat pumps amounts to almost nine million tonnes

  9. Small scale power production

    Energy Technology Data Exchange (ETDEWEB)

    Muoniovaara, M [IVO International Ltd, Vantaa (Finland)

    1997-12-31

    IVO International is a major constructor of biomass power plants in Finland and abroad. As a subsidiary of Imatran Voima Oy, the largest power utility in Finland, it has designed and constructed ten power plants owned by IVO Group or others capable of burning biomasses. Sizes of the plants vary from the world`s largest condensing peat-fired power plant of 155 MWe to a 6 MWe combined heat and power producing unit. This article describes the biomass power plants designed and constructed by IVO Group 3 refs.

  10. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

    1997-01-01

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  11. Use of experience for the improvement of technical specifications for operation

    International Nuclear Information System (INIS)

    Schweitz, J.P.; Seveon, J.J.

    1987-11-01

    The lessons drawn from experience led EdF to define in 1980 a national specification standard, to be applied to the first standardized series of 900 MWe units and to be adapted to the later 1300 MWe units. This document is periodically revised, every two years or so, to take into account new trends in safety thinking and lessons learnt from operating experience. In this paper we present the main developments in the field of technical specifications, particularly: the definition of rules applicable in the event of unavailability of safety-related equipment, the integration of plant unit operating feedback for specifications optimization

  12. Radioactive waste processing

    International Nuclear Information System (INIS)

    Dejonghe, P.

    1978-01-01

    This article gives an outline of the present situation, from a Belgian standpoint, in the field of the radioactive wastes processing. It estimates the annual quantity of various radioactive waste produced per 1000 MW(e) PWR installed from the ore mining till reprocessing of irradiated fuels. The methods of treatment concentration, fixation, final storable forms for liquid and solid waste of low activity and for high level activity waste. The storage of radioactive waste and the plutonium-bearing waste treatement are also considered. The estimated quantity of wastes produced for 5450 MW(e) in Belgium and their destination are presented. (A.F.)

  13. Small scale power production

    Energy Technology Data Exchange (ETDEWEB)

    Muoniovaara, M. [IVO International Ltd, Vantaa (Finland)

    1996-12-31

    IVO International is a major constructor of biomass power plants in Finland and abroad. As a subsidiary of Imatran Voima Oy, the largest power utility in Finland, it has designed and constructed ten power plants owned by IVO Group or others capable of burning biomasses. Sizes of the plants vary from the world`s largest condensing peat-fired power plant of 155 MWe to a 6 MWe combined heat and power producing unit. This article describes the biomass power plants designed and constructed by IVO Group 3 refs.

  14. Qualification of PHT piping of Indian 500 MW PHWR for LBB, using R-6 method

    International Nuclear Information System (INIS)

    Rastogi, Rohit; Bhasin, V.; Kushwaha, H.S.

    1997-01-01

    This document discusses the qualification of straight pipe portion of the primary heat transport (PHT) piping of Indian 500 MWe pressurised heavy water reactor (PHWR) for leak before break (LBB). The evaluation is done using R-6 [1] method. The results presented here are: the safety margins which exist on straight pipe components of main PHT piping of 500 MWe, under leakage size crack (LSC) and design basis accident loads; the sensitivity of safety margins with respect to different analysis parameters and the qualification of PHT piping for LBB based on criterion given by NUREG-1061 [2] and TECDOC-774 [3]. (author)

  15. Optimization of technology and boiler control to improve economical and environmental parameters

    Energy Technology Data Exchange (ETDEWEB)

    Stosek, V.; Neuman, P.; Mechura, V.; Masek, Z. [EGU Prague (Czechoslovakia)

    1995-12-01

    For cutting emissions NO{sub x} and CO in the Czech Republic are mostly applied primary measurers. At the same time measuring and control systems are innovated. Analog control systems are replaced by digital and computer network is developed in the power energy generation. It enables application of sophisticated information and diagnostic systems. It is shown how the EGU designs modification of technology equipment, measurement and control systems to increase efficiency and cut NO{sub x} emission levels at 110 MWe units at Prunerov power station and 200 MWe units at Tusimice before and after reconstruction are presented.

  16. Solar central receiver hybrid power system, Phase I. Volume 3. Appendices. Final technical report, October 1978-August 1979

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-09-01

    A design study for a central receiver/fossil fuel hybrid power system using molten salts for heat transfer and heat storage is presented. This volume contains the appendices: (A) parametric salt piping data; (B) sample heat exchanger calculations; (C) salt chemistry and salt/materials compatibility evaluation; (D) heliostat field coordinates; (E) data lists; (F) STEAEC program input data; (G) hybrid receiver design drawings; (H) hybrid receiver absorber tube thermal math model; (I) piping stress analysis; (J) 100-MWe 18-hour storage solar central receiver hybrid power system capital cost worksheets; and (K) 500-MWe 18-hour solar central receiver hybrid power system cost breakdown. (WHK)

  17. Factors affecting the minimum capital cost of a tokamak reactor

    International Nuclear Information System (INIS)

    Hancox, R.

    1981-01-01

    The Mk IIA Culham conceptual tokamak reactor design is a 2500 MWe steady-state reactor developed on the basis of a cost optimisation. A revised 1200 MWe conceptual design, the Mk IIB, used a lower wall loading and lower thermodynamic efficiency. A detailed costing of the Mk IIB design, however, showed it to have an unacceptably high capital cost. Since this high cost is a common characteristic of many fusion reactor designs, the cost optimisation of the Mk II design has been reconsidered. (author)

  18. Safety experience on EDF's PWRs

    International Nuclear Information System (INIS)

    Tanguy, P.

    1986-01-01

    The french nuclear programme has been widely publicized. In 1985, the total nuclear electricity generated was around 216 GWh, i. e. 70% of the electricity produced by electricity de France (EDF). If we consider only pressurized water reactors, at the end of 1985, 37 units were in operation (32 900 MWe and 5 1300 MWe) and 18 were under construction. I intend to review our experience with the safety of PWR's, but I will first present briefly some aspects related to the safety organization in France and the standardization policy. (author) [pt

  19. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  20. Nuclear desalination activities in India

    International Nuclear Information System (INIS)

    Bhattacharjee, B.

    1999-01-01

    The main emphasis of this article is on utilization of nuclear energy for desalination. Nuclear desalination is cheaper, eco-friendly and assists in sustainable growth of total energy generation programme in a country. PHWR type reactors are the main stay of nuclear energy programme in India. Nuclear waste heat for desalination is available in the moderator system of the 220 MW(e) and 500 MW(e) PHWRs. The low temperature evaporation technology (LET) for producing pure water from sea water is also discussed

  1. Review of the activities in Japan

    International Nuclear Information System (INIS)

    Otake, I.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress through.operation of the experimental fast reactor JOYO, design of the prototype fast breeder reactor MONJU and related R and D works. JOYO began operation in mid-1977, increased power from 50 MWt to 75 MWth in July 1979 and operation cycles at 75 MWth are continued at present. With respect to MONJU, which is a 300 MWe plant, progress toward construction has been made and the safety review are started by the concerned authorities. Conceptual design studies of large demonstration fast breeder reactor are also being made by PNC. It is a 1000 MWe, loop type plant

  2. Condition monitoring of primary coolant pump-motor units of Indian PHWR

    International Nuclear Information System (INIS)

    Rshikesan, P.B.; Sharma, S.S.; Mhetre, S.G.

    1994-01-01

    As the primary coolant pump motor units are located in shut down accessible area, their start up, satisfactory operation and shut down are monitored from control room. As unavailability of one pump in standardised 220 MWe station reduces the station power to about 110 MWe, satisfactory operation of the pump is also important from economic considerations. All the critical parameters of pump shaft, mechanical seal, bearing system, motor winding and shaft displacement (vibrations) are monitored/recorded to ensure satisfactory operation of critical, capital intensive pump-motor units. (author). 2 tabs., 1 fig

  3. Contributions of fast breeder test reactor to the advanced technology in India

    International Nuclear Information System (INIS)

    Kapoor, R.P.

    2001-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe loop type, sodium cooled, plutonium rich mixed carbide fuelled reactor. Its operation at Indira Gandhi Centre for Atomic Research, since first criticality in 1985, has contributed immensely to the advancement of this multidisciplinary and complex fast breeder technology in the country. It has also given a valuable operational feedback for the design of 500 MWe Prototype Fast Breeder Reactor. This paper highlights FBTR's significant contributions to this important technology which has a potential to provide energy security to the country in future. (author)

  4. Ontario Hydro decontamination experience

    International Nuclear Information System (INIS)

    Lacy, C.S.; Patterson, R.W.; Upton, M.S.

    1991-01-01

    Ontario Hydro currently operates 18 nuclear electric generating units of the CANDU design with a net capacity of 12,402 MW(e). An additional 1,762 MW(e) is under construction. The operation of these facilities has underlined the need to have decontamination capability both to reduce radiation fields, as well as to control and reduce contamination during component maintenance. This paper presents Ontario Hydro decontamination experience in two key areas - full heat transport decontamination to reduce system radiation fields, and component decontamination to reduce loose contamination particularly as practised in maintenance and decontamination centres. (author)

  5. Research and development of CO2 Capture and Storage Technologies in Fossil Fuel Power Plants

    Directory of Open Access Journals (Sweden)

    Lukáš Pilař

    2012-01-01

    Full Text Available This paper presents the results of a research project on the suitability of post-combustion CCS technology in the Czech Republic. It describes the ammonia CO2 separation method and its advantages and disadvantages. The paper evaluates its impact on the recent technology of a 250 MWe lignite coal fired power plant. The main result is a decrease in electric efficiency by 11 percentage points, a decrease in net electricity production by 62 MWe, and an increase in the amount of waste water. In addition, more consumables are needed.

  6. Three-dimensional model of the thermo-hydrodynamic neutron interaction in the core of water reactors (stationary states)

    International Nuclear Information System (INIS)

    Mastrangelo, Victor.

    1977-01-01

    A thermo-hydrodynamic neutron interaction model for permanent working conditions is developed in the case of closed circuits (boiling water reactors) and open circuits (pressurized water reactors). Two numerical convergence acceleration methods are then worked out for the resolution of linear problems by successive iterations. A physical study is devoted to the convergence of the thermo-hydrodynamic neutron interaction process. The model developed is applied to the calculation of the power distribution for the core of a 980 MWe BWR-6 type boiling water power station and to the study of normal and accidental working configurations of the pressurized water core of a 900 MWe PWR-CP1 unit [fr

  7. The GBR reactor an economically competitive Breeder

    International Nuclear Information System (INIS)

    Chermanne, J.

    1974-01-01

    In this article the design is described of a 1200 MWe fast breeder, gas-cooled reactor (GBR-4), prepared by a group of experts of the Gas Breeder Reactor Association and used as a reference system for economical and safety evaluations, as well as for defining the research and development program focussed on such concept and the specifications of the prospective demonstrative plant

  8. Modelberekening naar de invloed van lokale emissiebronnen van luchtverontreinigende componenten op de lokale vorming van fotochemische smog. Modellering van een (pluim)rookgasverspreidingsmodel, waaraan een beperkte subroutine met fotochemische en chemische reacties is toegevoegd

    NARCIS (Netherlands)

    van Rossum GJ; Erbrink JJ; de Leeuw FAAM

    1993-01-01

    The contribution of a 250 MWe co-generation plant assumed to be located in an urban area with about 300,000 inhabitants, to the photochemical ozone formation on the local scale is estimated by means of the flue gas dispersion model STACKS. In this study a limited number of photochemical reactions

  9. Pramana – Journal of Physics | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics. G PANDIKUMAR. Articles written in Pramana – Journal of Physics. Volume 90 Issue 4 April 2018 pp 46 Research Article. Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF/B-VII.1.

  10. The leaching of radioactivity from highly radioactive glass blocks buried below the water table: fifteen years of results

    International Nuclear Information System (INIS)

    Merritt, W.F.

    1976-03-01

    The results from two test burials of high-level fission products incorporated into nepheline syenite glass indicate that the nuclear wastes from fuel processing for a 30,000 MWe nuclear power industry could be incorporated into such glass and stored beneath the water table in the waste management area of Chalk River Nuclear Laboratories (CRNL) without harm to the environment. (author)

  11. DEMONSTRATION OF SORBENT INJECTION TECHNOLOGY ON A TANGENTIALLY COAL-FIRED UTILITY BOILER (YORKTOWN LIMB DEMONSTRATION)

    Science.gov (United States)

    The report summarizes activities conducted and results achieved in an EPA-sponsored program to demonstrate Limestone Injection Multistage Burner (LIMB) technology on a tangentially fired coal-burning utility boiler, Virginia Power's 180-MWe Yorktown Unit No. 2. his successfully d...

  12. H. B. Robinson Plant, Unit 2. Semiannual operating report No. 11, July--December 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 2,119,115 MW(e) with the generator on line 3,308 hrs. Information is presented concerning operation, power generation, shutdowns, corrective maintenance, occupational radiation exposure, release of radioactive materials, tests, inspections, refueling, and steam generator outage

  13. CO2 capture from power plants: Part I. A parametric study of the technical performance based on monoethanolamine

    NARCIS (Netherlands)

    Abu-Zahra, Mohammad R.M.; Schneiders, Léon H.J.; Niederer, John; Feron, Paul H.M.; Versteeg, Geert

    2007-01-01

    Capture and storage of CO2 from fossil fuel fired power plants is drawing increasing interest as a potential method for the control of greenhouse gas emissions. An optimization and technical parameter study for a CO2 capture process from flue gas of a 600 MWe bituminous coal fired power plant, based

  14. Assessment of antioxidant indices after incorporating crude oil ...

    African Journals Online (AJOL)

    They were divided into six groups of five rats each as follows: group 1: control, group 2: rats fed crude petroleum oil contaminated catfish diet (CPO-CCD) only, group 3: ... Administration of MWE, MEE and MDEE to the rats fed CPO-CCD significantly (p<0.05) increased the level of blood GSH, blood GSSG, SOD, CAT and ...

  15. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  16. Reorganization of AECL and the future marketing program

    International Nuclear Information System (INIS)

    Donnelly, James

    Atomic Energy of Canada Ltd. Engineering Co. has been reorganized to support the new emphasis on foreign sales of CANDU reactors. Much has been learned from reactor sales to Argentina, Korea, and Romania, but Canada needs to sell one 600 MWe reactor a year in order to avoid a decline in its nuclear industry. (LL)

  17. Agreement of 27 September 1988 between the International Atomic Energy Agency and the Government of India for the application of safeguards in connection with the supply of a nuclear power station from the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    1989-01-01

    The full text of the Agreement of 27 September 1988 between the Agency and the Government of India for the application of safeguards in connection with the supply of a nuclear power station composed of two pressurized light water reactors, each of 1000 MW(e) from the Union of Soviet Socialist Republics is reproduced

  18. Nuclear power perspective in China

    International Nuclear Information System (INIS)

    Liu Xinrong; Xu Changhua

    2003-01-01

    China started developing nuclear technology for power generation in the 1970s. A substantial step toward building nuclear power plants was taken as the beginning of 1980 s. The successful constructions and operations of Qinshan - 1 NPP, which was an indigenous PWR design with the capacity of 300 MWe, and Daya Bay NPP, which was an imported twin-unit PWR plant from France with the capacity of 900 MWe each, give impetus to further Chinese nuclear power development. Now there are 8 units with the total capacity of 6100 MWe in operation and 3 units with the total capacity of 2600 MWe under construction. For the sake of meeting the increasing demand for electricity for the sustainable economic development, changing the energy mix and mitigating the environment pollution impact caused by fossil fuel power plant, a near and middle term electrical power development program will be established soon. It is preliminarily predicted that the total power installation capacity will be 750-800GWe by the year 2020. The nuclear share will account for at least 4.0-4.5 percent of the total. This situation leaves the Chinese nuclear power industry with a good opportunity but also a great challenge. A practical nuclear power program and a consistent policy and strategy for future nuclear power development will be carefully prepared and implemented so as to maintain the nuclear power industry to be healthfully developed. (author)

  19. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  20. Tandem mirror reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Barr, W.L.; Carlson, G.A.

    1977-01-01

    A parametric analysis and a preliminary conceptual design for a 1000 MWe Tandem Mirror Reactor (TMR) are described. The concept is sufficiently attractive to encourage further work, both for a pure fusion TMR and a low technology TMR Fusion-Fission Hybrid