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Sample records for radially heterogeneous 1000-mwe

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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)

  6. '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

  7. 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

  8. 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

  9. 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

  10. European utility requirements (EUR) volume 3 assessment for AP1000

    International Nuclear Information System (INIS)

    Saiu, G.; Demetri, K.J.

    2005-01-01

    The EUR (European Utility Requirements) Volume 3 is intended to report the Plant Description, the Compliance Assessment to EUR Volumes 1 and 2, and finally, the Specific Requirements for each specific Nuclear Power Plant Design considered by the EUR. Five subsets of EUR Volume 3, based on EUR Revision B, are already published; all of which are next generation plant designs being developed for Europe beyond 2000. They include : 1) EP1000 - Passive Pressurized Light Water Reactor (3-Loop, 1000 MWe) 2) EPR - Evolutionary Pressurized Light Water Reactor (1500 MWe) 3) BWR90/90+ - Evolutionary Boiling Water Reactor (1400 MWe) 4) ABWR - Evolutionary Boiling Water Reactor (1400 MWe) 5) SWR 1000 - Boiling Water Reactor With Passive Features (1000 MWe) In addition, the following subsets are currently being developed: 1) AP1000 - Passive Pressurized Light Water Reactor (2-Loop, 1117 MWe) 2) VVER AES 92 - Pressurized Water Reactor With Passive Features (1000 MWe) The purpose of this paper is to provide an overview of the program, which started in January 2004 with the EUR group to prepare an EUR Volume 3 Subset for the AP1000 nuclear plant design. The AP1000 EUR compliance assessment, to be performed against EUR Revision C requirements, is an important step for the evaluation of the AP1000 design for application in Europe. The AP1000 compliance assessment is making full use of AP1000 licensing documentation, EPP Phase 2 design activities and EP1000 EUR detailed compliance assessment. As of today, nearly all of the EUR Chapters have been discussed within the EUR Coordination Group. Based on the results of the compliance assessment, it can be stated that the AP1000 design shows a good level of compliance with the EUR Revision C requirements. Nevertheless, the compliance assessment has highlighted areas for where the AP1000 plant deviates from the EUR. The EPP design group has selected the most significant ones for performing detailed studies to quantify the degree of compliance

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. Heterogeneous LTE-Advanced Network Expansion for 1000x Capacity

    DEFF Research Database (Denmark)

    Hu, Liang; Sanchez, Maria Laura Luque; Maternia, Michal

    2013-01-01

    this paper studies LTE (Long-Term Evolution)-Advanced heterogeneous network expansion in a dense urban environment for a 1000 times capacity increase and a 10 times increase in minimum user data rate requirements. The radio network capacity enhancement via outdoor and indoor small cell densificat...

  16. Steam generator design requirements for ACR-1000

    International Nuclear Information System (INIS)

    Subash, S.; Hau, K.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) has developed the ACR-1000 (Advanced CANDU Reactor-1000 ) to meet market expectations for enhanced safety of plant operation, high capacity factor, low operating cost, increased operating life, simple component replacement, reduced capital cost, and shorter construction schedule. The ACR-1000 design is based on the use of horizontal fuel channels surrounded by a heavy water moderator, the same feature as in all CANDU reactors. The major innovation in the ACR-1000 is the use of low enriched uranium fuel, and light water as the coolant, which circulates in the fuel channels. This results in a compact reactor core design and a reduction of heavy water inventory, both contributing to a significant decrease in capital cost per MWe produced. The ACR-1000 plant is a two-unit, integrated plant with each unit having a nominal gross output of about 1165 MWe with a net output of approximately 1085 MWe. The plant design is adaptable to a single unit configuration, if required. This paper focuses on the technical considerations that went into developing some of the important design requirements for the steam generators for the ACR-1000 plant and how these requirements are specified in the Technical Specification, which is the governing document for the steam generator (SG) detail design. Layout of these SGs in the plant is briefly described and their impacts on the SG design. (author)

  17. AP1000: Meeting economic goals in a competitive world. Annex 7

    International Nuclear Information System (INIS)

    Davis, G.; Cummins, E.; Winters, J.

    2002-01-01

    In the U.S., conditions are becoming more favorable for considering the nuclear option again for new baseload generation. While oil and natural gas prices have risen, the cost of operating the existing fleet of nuclear plants has decreased. Furthermore, an advanced 1000 MWe nuclear plant that will be even more cost-competitive with fossil fuels and natural gas will be available by 2005. Westinghouse, in an effort to further improve on the AP600's cost competitiveness, has developed the AP1000, a two-loop, 1000 MWe, advanced pressurized water reactor (PWR) with passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. Like the AP600, the AP1000 uses proven technology that builds on over 30 years of operating PWR experience. Westinghouse has completed design studies that demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MWe, maintaining its current design configuration and licensing basis. To maximize the cost savings, the AP1000 has been designed within the space constraints of the AP600, while retaining the credibility of proven components and substantial safety margins. The affect on the plant's overnight cost of the increased major components that is required to uprate the AP600 to 1000 MWe is small. This overall cost addition is on the order of 11 percent, while the overall power increase is almost 80 percent. This paper describes the changes made to uprate the AP600 and gives an overview of the plant design. (author)

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. A Probability Analysis of the Generating Cost for APR1000+

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Gag-Hyeon; Kim, Dae-Hun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO{sub 2} emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000+ development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000+(1,000MWe, 3.5 generation) will be developed based on APR+ (1,500MWe, 3.5 generation). And comparing to OPR1000(Korean Standard Nuclear power Plant, 2.5 generation), APR1000+ have many design features such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. In this simulation, the results of generating cost for APR1000+ preliminary conceptual design using a probability method was shown to be 48.37 ~ 74.22 won/kWh(median value 56.51 won/kWh). Those of OPR1000 was 42.08 ~ 61.77 won/kWh(median value 48.63 won/kWh). APR1000+ has -16.2% cost advantage over OPR1000 nuclear power plant. The main reason of this results is due to adding several safety designs.

  3. Key developments of the EP1000 design

    International Nuclear Information System (INIS)

    Noviello, L.

    1999-01-01

    In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The Phase I of the European Passive Plant program involved the evaluation of the Westinghouse 600 MWe AP600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs against the European Utility Requirements (EUR), and when necessary, the investigation of possible modifications to achieve compliance with the EUR. In Phase 1 of the program, which has been completed in 1996, the following major tasks were accomplished: The impacts of the European Utility Requirements (EUR) on the Westinghouse nuclear island design were evaluated. A 1000 MWe passive plant reference design (EP1000) was developed which conforms to the EUR and is expected to be licensable in Europe. With respect to the NSSS and containment, the EP1000 reference design closely follows those of the Westinghouse SPWR design, while the AP600 design has been taken as the basis for the design of the auxiliary systems. Extensive design and testing efforts have been made for the AP600 and SPWR during the respective multi-year programs. While the results of these programs have been and will continue to be utilised, at the maximum extent, to minimise the work to be performed on the EP1000 design, the compliance with EUR is a key design requirement for the EP1000 The ultimate objective of Phase 2 of the program is to develop design details and perform supporting analyses to produce a Safety Case Report (SCR) for submittal to European Safety Authorities. The first part of Phase 2, hereafter referred as Phase 2A, started at the beginning of 1997 and will be completed at the end of 1998. Scope of this phase of the program is to develop the design modifications of important systems and structures so to comply with the

  4. 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)

  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. 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)

  7. EP 1000 -The European Passive Plant

    International Nuclear Information System (INIS)

    Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

    1998-01-01

    A group of European utilities, along with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT) initiated a program to evaluate Westinghouse passive nuclear plant technology for application in Europe. The European utility group consisted of: Agrupacion electrica para al Desarrollo Technologico Nuclear (DTN), Spain; Electricite de France; ENEL, SpA., Italy; IVO Power Engineering, Ltd., Finland; Scottish Nuclear Limited (acting for itself on behalf of Nuclear Electric plc, U.K.; Tractebel Energy Engineering, Belgium; UAK (represented by NOK-Beznau), Switzerland; and Vattenfall AB, Ringhals, Sweden. The European Passive Plant (EPP) program, which began in 1994, is an evaluation of the Westinghouse 600 MWe AP 600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs in meeting the European Utility Requirements (EUR), and where necessary, modifying the design to achieve compliance. Phase 1 or the EPP program was completed and included the two major tasks of evaluating the effect of the EUR on the Westinghouse nuclear island and developing the EP 1000, a 1000 MWe passive plant reference design that conforms to the EUR and would be licensable in Europe. The EP 1000 closely follows the Westinghouse SPWR design for safety systems and containment and the AP 600 design for auxiliary systems. It also includes features that where required to meet the EUR and key European licensing requirements. The primary circuit of the EP 1000 retains most of the general features of the current-day designs, but some evolutionary features to enhance reliability, simplicity of operation, ease of maintenance, and plant safety have been incorporated into the design. The core, reactor vessel, and reactor internals of the EP 1000 are similar to those of currently operating Westinghouse PWR plants, but several new features are included to enhance the performance characteristics. The basic EP 1000 safety philosophy is based on use of inherent

  8. 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.)

  9. 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)

  10. 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

  11. VVER-1000 backfitting programs

    International Nuclear Information System (INIS)

    Zabka, H.; Milhem, J.L.

    1998-01-01

    Russia, Ukraine, and Bulgaria have nineteen nuclear generating units of the VVER-1000/V-320 (1000 MWe PWR) type in operation. Most of these plants were built in the eighties. Their design is based on Soviet standards of the seventies. In the early eighties and, in particular, after the Chernobyl accident, new safety principles and supplementary specific standards were introduced. However, they were taken into account only to a limited extent in the design and construction of the VVER-1000/V-320 plants. A number of nuclear power plants, whose construction was stopped after the political changes in the countries of the former USSR, now are to be completed with the financial assistance of the Commission of the European Union and other Western organizations, respectively. This Western support is dependent on the condition that these plants attain a level of engineered safeguards comparable to that of PWR plants currently in operation in Western Europe. (orig.) [de

  12. 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)

  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. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. Probability Analysis of the Construction Cost of an APR1000 Single Unit

    International Nuclear Information System (INIS)

    Ha, Gak Hyeon; Suh, Yong Pyo; Kang, Yong Chul

    2011-01-01

    The nuclear power plant market is expected to grow rapidly in order to address issues of global warming, cutting CO 2 emissions and securing stable electricity supplies. Under these circumstances, the main primary goal of the APR1000 development is to ensure export competitiveness in the developing countries in the Middle East and Southeast Asia. To that end, APR1000 (1,000MWe, 3 rd generation) will be developed based on the OPR1000 (Korean standard nuclear power plant, 2.5 generation) by incorporating and improving the general requirements such as the 60 year design life time, comprehensive site requirement of 0.3g seismic design, stability improvement, operability improvement and provisions for severe accidents. The APR1000 adds 16 advanced design features to its predecessor, as outlined below in Table 1

  1. 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

  2. Westinghouse AP1000 licensing maturity

    International Nuclear Information System (INIS)

    Schulz, T.; Vijuk, R.P.

    2005-01-01

    The Westinghouse AP1000 Program is aimed at making available a nuclear power plant that is economical in the U.S deregulated electrical power industry in the near-term. The AP1000 is two-loop 1000 MWe pressurizer water reactor (PWR). It is an up rated version of the AP600. The AP1000 uses passive safety systems to provide significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 35 years of operating PWR experience. The AP1000 received Final Design Approval by the United States Nuclear Regulatory Commission (U.S. NRC) in September 2004. The AP1000 meets the US utility requirements. The AP1000 and its sister plant the AP600 have gone through a very through and complete licensing review. This paper describes the U.S. NRC review efforts of both the AP600 and the AP1000. The detail of the review and the independent calculations, evaluations and testing is discussed. The AP600 licensing documentation was submitted in 1992. The U.S. NRC granted Final Design Approval in 1999. During the intervening 7 years, the U.S. NRC asked thousands of questions, performed independent safety analysis, audited Westinghouse calculations and analysis, and performed independent testing. The more significant areas of discussion will be described. For the AP1000 Westinghouse first engaged the U.S. NRC in pre-certification discussions to define the extent of the review required, since the design is so similar to the AP600. The AP1000 licensing documentation was submitted in March 2002. The U.S. NRC granted Final Design Approval in September 2004. During the intervening 2 1/2 years, the U.S. NRC asked hundreds of questions, performed independent safety analysis, audited Westinghouse calculations and analysis, and performed independent testing. The more significant areas of discussion will be described. The implications of this review and approval on AP1000 applications in

  3. 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.

  4. The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

    Science.gov (United States)

    Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

    2018-02-01

    The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

  5. Finite element modeling of AP1000 nuclear island

    International Nuclear Information System (INIS)

    Tinic, S.; Orr, R.

    2003-01-01

    The AP1000 is a standard design developed by Westinghouse and its partners for an advanced nuclear power plant utilizing passive safety features. It is based on the certified design of the AP600 and has been uprated to 1000 MWe. The plant has five principal building structures; the nuclear island, the turbine building; the annex building; the diesel generator building and the radwaste building. The nuclear island consists of the containment building (the steel containment vessel and the containment internal structures), the shield building, and the auxiliary building. These structures are founded on a common basemat and are collectively known as the nuclear island. This paper describes use of the general purpose finite element program ANSYS [2] in structural analyses and qualification of the AP1000 nuclear island buildings. It describes the modeling of the shield building and the auxiliary building and the series of analyses and the flow of information from the global analyses to the detailed analyses and building qualification. (author)

  6. 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

  7. Dynamical links between small- and large-scale mantle heterogeneity: Seismological evidence

    Science.gov (United States)

    Frost, Daniel A.; Garnero, Edward J.; Rost, Sebastian

    2018-01-01

    We identify PKP • PKP scattered waves (also known as P‧ •P‧) from earthquakes recorded at small-aperture seismic arrays at distances less than 65°. P‧ •P‧ energy travels as a PKP wave through the core, up into the mantle, then scatters back down through the core to the receiver as a second PKP. P‧ •P‧ waves are unique in that they allow scattering heterogeneities throughout the mantle to be imaged. We use array-processing methods to amplify low amplitude, coherent scattered energy signals and resolve their incoming direction. We deterministically map scattering heterogeneity locations from the core-mantle boundary to the surface. We use an extensive dataset with sensitivity to a large volume of the mantle and a location method allowing us to resolve and map more heterogeneities than have previously been possible, representing a significant increase in our understanding of small-scale structure within the mantle. Our results demonstrate that the distribution of scattering heterogeneities varies both radially and laterally. Scattering is most abundant in the uppermost and lowermost mantle, and a minimum in the mid-mantle, resembling the radial distribution of tomographically derived whole-mantle velocity heterogeneity. We investigate the spatial correlation of scattering heterogeneities with large-scale tomographic velocities, lateral velocity gradients, the locations of deep-seated hotspots and subducted slabs. In the lowermost 1500 km of the mantle, small-scale heterogeneities correlate with regions of low seismic velocity, high lateral seismic gradient, and proximity to hotspots. In the upper 1000 km of the mantle there is no significant correlation between scattering heterogeneity location and subducted slabs. Between 600 and 900 km depth, scattering heterogeneities are more common in the regions most remote from slabs, and close to hotspots. Scattering heterogeneities show an affinity for regions close to slabs within the upper 200 km of the

  8. 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)

  9. 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

  10. 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)

  11. Learning through delivery, Westinghouse AP1000 plant construction

    International Nuclear Information System (INIS)

    Gorgemans, J.; Hinman, R.D.; Steuck, C.M.; Greco, P.L.

    2014-01-01

    The AP1000 plant, which is a 1100 MWe class pressurized water reactor with passive safety features, is designed around a conventional 2 loop, 2 steam generator primary system configuration with 2 hot legs, 4 reactor coolant pumps directly mounted in the steam generator lower head and 4 cold legs. A particular feature of AP1000 is its modular construction to minimize the time and cost of construction. Modular construction allows activities to be run in parallel, it allows more activities to be performed in a controlled factory instead of in the field, and it provides a better level of quality. The AP1000 plant design includes 106 structural modules and 52 mechanical modules. Structural modules include all penetrations for piping, cable trays, HVAC duct runs, and all reinforcement for pipe, equipment hangers, and supports. Structural modules are shipped in sub-modules to support transportation by rail or truck or barge. Mechanical modules contain equipment such as pumps, tanks, heat exchangers, air-handling units, and filters along with interconnecting pipes, valves, instruments, wiring and support services. Modular construction requires strong coordination between engineering, supply chain and construction. A total of 8 AP1000 units are currently under construction in China and in the United States. The lessons learned and best practices of each new AP1000 construction are systematically incorporated into the standard design. (A.C.)

  12. 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)

  13. AP1000 plant construction in China: Ansaldo Nucleare contribution

    International Nuclear Information System (INIS)

    Frogheri, Monica; Saiu, Gianfranco

    2009-01-01

    On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract includes: the design of the steel containment vessel, preparation of construction and fabrication, specifications, design and supply of SCV mechanical penetrations, air locks and equipment hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China. (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. Design study on metal fuel FBR cores

    International Nuclear Information System (INIS)

    Yokoo, T.; Tanaka, Y.; Ogata, T.

    1991-01-01

    A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

  16. 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

  17. ACR-1000 design provisions for severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

  18. 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...

  19. 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

  20. 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

  1. 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)

  2. 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)

  3. Local heterogeneity effects on small-sample worths

    International Nuclear Information System (INIS)

    Schaefer, R.W.

    1986-01-01

    One of the parameters usually measured in a fast reactor critical assembly is the reactivity associated with inserting a small sample of a material into the core (sample worth). Local heterogeneities introduced by the worth measurement techniques can have a significant effect on the sample worth. Unfortunately, the capability is lacking to model some of the heterogeneity effects associated with the experimental technique traditionally used at ANL (the radial tube technique). It has been suggested that these effects could account for a large portion of what remains of the longstanding central worth discrepancy. The purpose of this paper is to describe a large body of experimental data - most of which has never been reported - that shows the effect of radial tube-related local heterogeneities

  4. 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

  5. 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.)

  6. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

    2006-01-01

    Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

  7. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  8. Confirmation of a change in the global shear velocity pattern at around 1000 km depth

    Science.gov (United States)

    Durand, S.; Debayle, E.; Ricard, Y.; Zaroli, C.; Lambotte, S.

    2017-12-01

    In this study, we confirm the existence of a change in the shear velocity spectrum around 1000 km depth based on a new shear velocity tomographic model of the Earth's mantle, SEISGLOB2. This model is based on Rayleigh surface wave phase velocities, self- and cross-coupling structure coefficients of spheroidal normal modes and body wave traveltimes which are, for the first time, combined in a tomographic inversion. SEISGLOB2 is developed up to spherical harmonic degree 40 and in 21 radial spline functions. The spectrum of SEISGLOB2 is the flattest (i.e. richest in 'short' wavelengths corresponding to spherical harmonic degrees greater than 10) around 1000 km depth and this flattening occurs between 670 and 1500 km depth. We also confirm various changes in the continuity of slabs and mantle plumes all around 1000 km depth where we also observed the upper boundary of Large Low Shear Velocity Provinces. The existence of a flatter spectrum, richer in short-wavelength heterogeneities, in a region of the mid-mantle can have great impacts on our understanding of the mantle dynamics and should thus be better understood in the future. Although a viscosity increase, a phase change or a compositional change can all concur to induce this change of pattern, its precise origin is still very uncertain.

  9. EP1000 passive plant description

    International Nuclear Information System (INIS)

    Saiu, G.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In Phase I of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) will be completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. The second part, 'Phase 2B', includes both the analyses and evaluations required to demonstrate the adequacy of the design, and to support the preparation of Safety Case Report. The second part of Phase 2 of the program will start at the beginning of 1999 and will be completed in the 2001. Incorporation of the EUR has been a key design requirement for the EP1000 from the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. This paper integrates and updates the plant description reported in the IAEA TECDOC-968. The most significant developments of the EP1000 plant design during Phase 2A of the EPP program are described and reference is made to the key design requirements set by the EUR Rev. B document. (author)

  10. 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)

  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. Effect of multicell DRAGON calculations depends on the environment on the DONJON predictions for the ACR-1000

    International Nuclear Information System (INIS)

    Duquette, J.-S.

    2009-01-01

    For understanding the behavior of a nuclear reactor core, it is necessary to make a full core calculation in order to compute the neutrons flux. To obtain the neutrons flux, solving the Boltzmann transport equation is required. That is not a simple task and it is impossible to analytically fend the solution of the neutrons transport equation on a complex core. Following a series of approximations, it is possible to numerically solve the neutrons transport equation. The solution of this equation is done step by step. Calculations will be performed over the ACR-1000 core. The Advanced CANDU Reactor (ACR-1000) is a generation III+ heavy water moderated and light water cooled reactor. It is a 1200 MW(e) power reactor. Amongst the ACR-1000 design parameters that differ from the CANDU 6, the reduced lattice pitch and the use of light water coolant and enriched fuel are the three most important. Those features modify the behavior of the neutrons in the ACR compared to the CANDU 6. The impact of the tight lattice is that a cell is more strongly coupled to its neighbor. The objective of this work is to determine the impact of the environment on the cell properties of the ACR-1000. Those properties will be used to perform full core calculations. The neutron transport calculations are performed with DRAGON whereas for the diffusion calculation on a full core. The code DONJON will be used. The DRAGON reference transport calculation will be made on a single cell. Then, a series of calculations will be performed using DRAGON over two types of assemblies, the first modelling the core interior and the second, modelling the core periphery. Moreover, the fuel age will sometimes be homogeneous, sometimes heterogeneous. The fuel will be burned during six hundred days. One thus obtains libraries of macroscopic cross sections over a six hundred days interval for various simulations. Thereafter, we will determine the effect of a neutrons transport multicell calculation on various DONJON

  13. 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

  14. Radial glial cells in the adult dentate gyrus: what are they and where do they come from?

    Science.gov (United States)

    Berg, Daniel A; Bond, Allison M; Ming, Guo-Li; Song, Hongjun

    2018-01-01

    Adult neurogenesis occurs in the dentate gyrus in the mammalian hippocampus. These new neurons arise from neural precursor cells named radial glia-like cells, which are situated in the subgranular zone of the dentate gyrus. Here, we review the emerging topic of precursor heterogeneity in the adult subgranular zone. We also discuss how this heterogeneity may be established during development and focus on the embryonic origin of the dentate gyrus and radial glia-like stem cells. Finally, we discuss recently developed single-cell techniques, which we believe will be critical to comprehensively investigate adult neural stem cell origin and heterogeneity.

  15. WWER radial reflector modeling by diffusion codes

    International Nuclear Information System (INIS)

    Petkov, P. T.; Mittag, S.

    2005-01-01

    The two commonly used approaches to describe the WWER radial reflectors in diffusion codes, by albedo on the core-reflector boundary and by a ring of diffusive assembly size nodes, are discussed. The advantages and disadvantages of the first approach are presented first, then the Koebke's equivalence theory is outlined and its implementation for the WWER radial reflectors is discussed. Results for the WWER-1000 reactor are presented. Then the boundary conditions on the outer reflector boundary are discussed. The possibility to divide the library into fuel assembly and reflector parts and to generate each library by a separate code package is discussed. Finally, the homogenization errors for rodded assemblies are presented and discussed (Author)

  16. Current status of generation III nuclear power and assessment of AP1000 developed by Westinghouse

    International Nuclear Information System (INIS)

    Zhang Mingchang

    2005-01-01

    In order to make greater contributions to the environment, new nuclear power systems will be needed to meet the increase of electricity demand and to replace plants to be decommissioned. A series of new designs, so called Generation III and Generation III +, are being developed to ensure their deployment in a Near-Term Deployment Road-map in US by 2010 and in Europe by 2015. The AP1000, developed by Westinghouse, is a two-loop 1000 MWe PWR with passive safety features and extensive simplifications to enhance its competitiveness in cost and tariff. It is the first Generation III + plant receiving the Final Design Approval by the US NRC. This paper briefly describes AP1000 design features and technical specifications, and presents a more detailed design evaluation with reference to relevant literatures. Both the opportunity and challenges for nuclear power development in China during the first decade of the 21 st century in a historic transition from Gen II to Gen III are analyzed. The key is to balance risks and benefits if the first AP1000 to be settled down in China. (author)

  17. 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.

  18. 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.

  19. Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

    International Nuclear Information System (INIS)

    Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

    2005-01-01

    Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

  20. Radial electron beam laser excitation: the REBLE report

    International Nuclear Information System (INIS)

    Ramirez, J.J.; Prestwich, K.R.

    1978-10-01

    The results of an investigation of techniques to generate high-power radially converging electron beams and the application of these beams to gas lasers is discussed. The design and performance of the REBLE accelerator that was developed for this program is presented. Reliable operation of the radial diode has been obtained at levels up to 1 MV, 200 kA, and 20 ns. It has been demonstrated that the anode current density can be made uniform to better than 15% over 1000 cm 2 areas with 100 to 250 A/cm 2 intensities. The measured total and spatially resolved energy deposition of this radial electron beam in various gases is compared with Monte Carlo calculations. In most cases, these codes give an accurate description of the beam transport and energy deposition. With the electron beam pumping xenon gas, the amplitude of xenon excimer radiation (1720 A 0 ) was radially uniform to within the experimental uncertainty. The efficiency of converting deposited electron beam energy to xenon excimer radiation was 20%

  1. 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)

  2. 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.)

  3. 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.)

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-01-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

  10. 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)

  11. 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)

  12. Radial Structure Scaffolds Convolution Patterns of Developing Cerebral Cortex

    Directory of Open Access Journals (Sweden)

    Mir Jalil Razavi

    2017-08-01

    Full Text Available Commonly-preserved radial convolution is a prominent characteristic of the mammalian cerebral cortex. Endeavors from multiple disciplines have been devoted for decades to explore the causes for this enigmatic structure. However, the underlying mechanisms that lead to consistent cortical convolution patterns still remain poorly understood. In this work, inspired by prior studies, we propose and evaluate a plausible theory that radial convolution during the early development of the brain is sculptured by radial structures consisting of radial glial cells (RGCs and maturing axons. Specifically, the regionally heterogeneous development and distribution of RGCs controlled by Trnp1 regulate the convex and concave convolution patterns (gyri and sulci in the radial direction, while the interplay of RGCs' effects on convolution and axons regulates the convex (gyral convolution patterns. This theory is assessed by observations and measurements in literature from multiple disciplines such as neurobiology, genetics, biomechanics, etc., at multiple scales to date. Particularly, this theory is further validated by multimodal imaging data analysis and computational simulations in this study. We offer a versatile and descriptive study model that can provide reasonable explanations of observations, experiments, and simulations of the characteristic mammalian cortical folding.

  13. 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)

  14. 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)

  15. 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)

  16. 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.)

  17. 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.

  18. Calculation of breaking radiation dose fields in heterogenous media by a method of the transformation of axial distribution

    International Nuclear Information System (INIS)

    Mil'shtejn, R.S.

    1988-01-01

    Analysis of dose fields in a heterogeneous tissue equivalent medium has shown that dose distributions have radial symmetry and can be described by a curve of axial distribution with renormalization of maximum ionization depth. A method of the calculation of a dose field in a heterogeneous medium using the principle of radial symmetry is presented

  19. 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)

  20. Analysis of radially heterogeneous ZPPR-13A benchmark for investigating the spatial dependence of the calculated-to-experiment ratio for control rod worths

    International Nuclear Information System (INIS)

    Mahalakshmi, B.; Mohanakrishnan, P.

    1993-01-01

    Investigation were performed on the ZPPR-13A critical assembly to determine the cause of the radial variation of the calculated-to-experimental (C/E) ratio for control rod worth in large heterogeneous cores. The effects of errors in cross section, mesh size, group condensation, transport, and modeling were studied by studied by using two- and three-dimensional diffusion calculations and three-dimensional transport calculations. In that process, the cross-section set and the calculation scheme that are being used for fast reactor design in India have been revalidated. The cross-section set was found to yield satisfactory results. Three-dimensional calculations with adjusted and unadjusted cross sections confirmed that the error in cross sections was largely responsible for the radial dependence of the C/E ratios. The contributions from group condensation and mesh size errors were < 2%, and from modeling errors and transport correction, < 1%. The effect of these errors is insignificant when compared with the effect of the cross-section error. The analysis also showed that even without the adjustment in diffusion coefficient suggested in earlier studies, a satisfactory prediction is found, at least for this benchmark. The diffusion-to-transport correction for control rod worth was found to be -7%

  1. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  2. 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.

  3. Effect of air condition on AP-1000 containment cooling performance in station black out accident

    International Nuclear Information System (INIS)

    Hendro Tjahjono

    2015-01-01

    AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

  4. 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)

  5. Radial glial cells in the adult dentate gyrus: what are they and where do they come from? [version 1; referees: 2 approved

    Directory of Open Access Journals (Sweden)

    Daniel A. Berg

    2018-03-01

    Full Text Available Adult neurogenesis occurs in the dentate gyrus in the mammalian hippocampus. These new neurons arise from neural precursor cells named radial glia-like cells, which are situated in the subgranular zone of the dentate gyrus. Here, we review the emerging topic of precursor heterogeneity in the adult subgranular zone. We also discuss how this heterogeneity may be established during development and focus on the embryonic origin of the dentate gyrus and radial glia-like stem cells. Finally, we discuss recently developed single-cell techniques, which we believe will be critical to comprehensively investigate adult neural stem cell origin and heterogeneity.

  6. 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

  7. A Bayesian approach to infer the radial distribution of temperature and anisotropy in the transition zone from seismic data

    Science.gov (United States)

    Drilleau, M.; Beucler, E.; Mocquet, A.; Verhoeven, O.; Moebs, G.; Burgos, G.; Montagner, J.

    2013-12-01

    Mineralogical transformations and matter transfers within the Earth's mantle make the 350-1000 km depth range (considered here as the mantle transition zone) highly heterogeneous and anisotropic. Most of the 3-D global tomographic models are anchored on small perturbations from 1-D models such as PREM, and are secondly interpreted in terms of temperature and composition distributions. However, the degree of heterogeneity in the transition zone can be strong enough so that the concept of a 1-D reference seismic model may be addressed. To avoid the use of any seismic reference model, we developed a Markov chain Monte Carlo algorithm to directly interpret surface wave dispersion curves in terms of temperature and radial anisotropy distributions, considering a given composition of the mantle. These interpretations are based on laboratory measurements of elastic moduli and Birch-Murnaghan equation of state. An originality of the algorithm is its ability to explore both smoothly varying models and first-order discontinuities, using C1-Bézier curves, which interpolate the randomly chosen values for depth, temperature and radial anisotropy. This parameterization is able to generate a self-adapting parameter space exploration while reducing the computing time. Using a Bayesian exploration, the probability distributions on temperature and anisotropy are governed by uncertainties on the data set. The method was successfully applied to both synthetic data and real dispersion curves. Surface wave measurements along the Vanuatu- California path suggest a strong anisotropy above 400 km depth which decreases below, and a monotonous temperature distribution between 350 and 1000 km depth. On the contrary, a negative shear wave anisotropy of about 2 % is found at the top of the transition zone below Eurasia. Considering compositions ranging from piclogite to pyrolite, the overall temperature profile and temperature gradient are higher for the continental path than for the oceanic

  8. AP1000 - the new standard for nuclear power

    International Nuclear Information System (INIS)

    Lipman, Daniel S.

    2006-01-01

    Full text of publication follows: The AP1000 is the only Generation III+ reactor to receive Final Design Approval (FDA) from the Nuclear Regulatory Commission, and is expected to receive its Design Certification by the end of the year. Building on the proven features of current generation nuclear plants, the AP1000 combines experience with innovation into a design that surpasses current standards of safety and reliability. Use of passive safety features results in a simpler and more compact design that enhances safety, simplifies O and M requirements, and reduces capital and operating costs. At 1117 Mwe, the AP1000 is well suited for almost any grid system and will be fully competitive with combined-cycle gas and comparable fossil fuel plants. The AP1000 is ready to help launch a renaissance in new nuclear plant construction throughout the world. Maturity of Design: In excess of 1300 man-years and $400 million in development funding have been expended on the AP1000. It has undergone extensive, part scale testing at the system, sub-system and component level, in addition to a series of part scale integrated tests. The AP1000 is the most analyzed of the next generation reactors. Simplicity of Design/Economics: The AP1000 uses simplified and innovative passive safety systems to an unprecedented extent. Simplified passive safety systems provide reliable operation, reduced capital costs, and enhanced plant safety with large plant operating margins. The AP1000 features improved reliability through simplicity rather than addition of redundant safety trains. This simpler design is easier and less costly to operate and maintain than larger, more complex plants, while less equipment and smaller buildings translate into lower capital costs and shorter construction durations. After construction, economic benefit will be found in reduced operating and maintenance costs, largely due to reduced operating and maintenance staffing requirements. Construction aspects

  9. Rupture of posterior cruciate ligament leads to radial displacement of the medial meniscus.

    Science.gov (United States)

    Zhang, Can; Deng, Zhenhan; Luo, Wei; Xiao, Wenfeng; Hu, Yihe; Liao, Zhan; Li, Kanghua; He, Hongbo

    2017-07-11

    To explore the association between the rupture of posterior cruciate ligament (PCL) and the radial displacement of medial meniscus under the conditions of different flexion and various axial loads. The radial displacement value of medial meniscus was measured for the specimens of normal adult knee joints, including 12 intact PCLs, 6 ruptures of the anterolateral bundle (ALB), 6 ruptures of the postmedial bundle (PMB), and 12 complete ruptures. The measurement was conducted at 0°, 30°, 60°, and 90° of knee flexion angles under 200 N, 400 N, 600 N, 800 N and 1000 N of axial loads respectively. The displacement values of medial meniscus of the ALB rupture group increased at 0° flexion under 800 N and 1000 N, and at 30°, 60° and 90° flexion under all loads in comparison with the PCL intact group. The displacement values of the PMB rupture group was higher at 0° and 90° flexion under all loads, and at 30° and 60° flexion under 800 N and 1000 N loads. The displacement of the PCL complete rupture group increased at all flexion angles under all loads. Either partial or complete rupture of the PCL can increase in the radial displacement of the medial meniscus, which may explain the degenerative changes that occuring in the medial meniscus due to PCL injury. Therefore, early reestablishment of the PCL is necessarily required in order to maintain stability of the knee joint after PCL injury.

  10. 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

  11. Transient well flow in vertically heterogeneous aquifers.

    NARCIS (Netherlands)

    Hemker, C.J.

    1999-01-01

    A solution for the general problem of computing well flow in vertically heterogeneous aquifers is found by an integration of both analytical and numerical techniques. The radial component of flow is treated analytically; the drawdown is a continuous function of the distance to the well. The

  12. 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

  13. 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)

  14. 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

  15. Neutronic study of nanofluids application to VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hadad, K., E-mail: hadad@email.arizona.ed [School of Engineering, Shiraz University, Shiraz 7134554115 (Iran, Islamic Republic of); Aerospace and Mechanical Engineering, University of Arizona, Tucson, AZ 85721 (United States); Hajizadeh, A.; Jafarpour, K. [School of Engineering, Shiraz University, Shiraz 7134554115 (Iran, Islamic Republic of); Ganapol, B.D. [Aerospace and Mechanical Engineering, University of Arizona, Tucson, AZ 85721 (United States)

    2010-11-15

    The change in neutronic parameters of the VVER-1000 nuclear reactor core attributable to the use of nanoparticle/water (nanofluid) as coolant is presented in this paper. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. We considered five nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper and Zirconia. The results of our study show that at low concentration (0.001 volume fraction) Alumina is optimum nanoparticle for normal operation. The maximum radial and axial LPPF were found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on fuel clad, a flux and K{sub eff} depression occurs and Al{sub 2}O{sub 3} has the lowest rate of drop off.

  16. 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.

  17. Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

    International Nuclear Information System (INIS)

    Carbajo, J.J.; Lindner, C.N.

    1992-01-01

    The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car

  18. 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

  19. 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

  20. Oxidation-induced deformation of zircaloy-4 tubing in steam in the temperature range 600-1000 degree C

    International Nuclear Information System (INIS)

    Aly, A.E.; Hussein, A.G.; EL-Raghy, S.M.; EL-Sayed, A.A.; EL-Banna, O.A.

    1992-01-01

    The oxidation-induced deformation of zircaloy-4 (zry-4) tubing in steam has been studied in the temperature range 600 to 1000 degree C. The induced deformation has been measured in both radial and axial directions of the tube. The effect of hydrogen addition to steam was also investigated. The oxidation-induced deformation has been characterized by uniform and non-uniform (distortion) strain period. During the uniform strain period the radial strain kinetics were found in general, to be parallel to the oxidation kinetics. The axial strain (δA) induced by oxidation was found to be always lower than the radial strain (εR). The addition of 5% by volume hydrogen to steam leads to an increase in the oxidation rate and to a decrease in the degree of anisotropy between radial and axial strains

  1. Development of 3-D FBR heterogeneous core calculation method based on characteristics method

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Maruyama, Manabu; Hamada, Yuzuru; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

    2002-01-01

    A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region

  2. 7 CFR 1000.91-1000.92 - [Reserved

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true [Reserved] 1000.91-1000.92 Section 1000.91-1000.92 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing... Miscellaneous Provisions §§ 1000.91-1000.92 [Reserved] ...

  3. Radial heterogeneity of some analytical columns used in high-performance liquid chromatography.

    Science.gov (United States)

    Abia, Jude A; Mriziq, Khaled S; Guiochon, Georges A

    2009-04-10

    An on-column electrochemical microdetector was used to determine accurately the radial distribution of the mobile phase velocity and of the column efficiency at the exit of three common analytical columns, namely a 100 mm x 4.6mm C18 bonded silica-based monolithic column, a 150 mm x 4.6mm column packed with 2.7 microm porous shell particles of C18 bonded silica (HALO), and a 150 mm x 4.6mm column packed with 3 microm fully porous C18 bonded silica particles (LUNA). The results obtained demonstrate that all three columns are not radially homogeneous. In all three cases, the efficiency was found to be lower in the wall region of the column than in its core region (the central core with a radius of 1/3 the column inner radius). The decrease in local efficiency from the core to the wall regions was lower in the case of the monolith (ca. 25%) than in that of the two particle-packed columns (ca. 35-50%). The mobile phase velocity was found to be ca. 1.5% higher in the wall than in the core region of the monolithic column while, in contrast, it was ca. 2.5-4.0% lower in the wall region for the two particle-packed columns.

  4. Heterogeneous recycling in SFR core periphery

    International Nuclear Information System (INIS)

    Varaine, Frederic; Buiron, Laurent; Boucher, Lionel; Chabert, Christine

    2008-01-01

    In the framework of next generation fast reactor design, the management of minor actinides (MA) is one of the key issues. The Transmutation of MA can be achieving with various modes of transmutation and waste management. Two ways for transmutation: - The homogeneous mode where the minor actinides to be transmuted are directly mixed with 'standard' fuel of the reactor, - The heterogeneous way for which the actinides to be transmuted are separated from the fuel itself, in limited number of S/A (targets) devoted to actinides transmutation. Associated with two ways for actinides management: - The multiple recycling: in this case whole or part of minor actinides and plutonium at the end of each reactor cycle is sent back in the following cycle. In that way, only reprocessing losses go to the waste, - The once-through way: in this case the minor actinides are transmuted in targets where very high burn up is reached. Fast reactors offer the best performances to transmute the minor actinides in homogeneous or heterogeneous way at industrial scale. The safety criteria are acceptable for all solutions if the MA content is not over 2.5% of the total heavy nuclides. In this context, the last results obtained for minor actinides transmutation in sodium fast reactor depleted uranium radial blankets are presented. This concept is based on a heterogeneous multiple recycling model. The use of the oxide matrix allows to reprocess such S/A in the spent fuel standard flow. For the study, we use a preliminary design of a 3600 MWth sodium Fast Reactor in progress at CEA. We investigate the transmutation performances of (U+Np+Am+Cm)O 2 fuel in radial blankets assemblies. We focus on two upper and lower assumptions in order to investigate the feasibility domain for this concept: one with a minor actinides (MA) content of 10%, and the second one with an enrichment of MA close to 40%. The CEA is studying scenarios of principle for the French case through a dynamic vision of the nuclear

  5. Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

    International Nuclear Information System (INIS)

    Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

    2011-01-01

    The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

  6. Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

    1997-03-01

    This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

  7. A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

  8. 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.

  9. 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%.

  10. Radial nerve dysfunction

    Science.gov (United States)

    Neuropathy - radial nerve; Radial nerve palsy; Mononeuropathy ... Damage to one nerve group, such as the radial nerve, is called mononeuropathy . Mononeuropathy means there is damage to a single nerve. Both ...

  11. 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

  12. 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)

  13. 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)

  14. Prenatal Diagnosis of Bilateral Ectrodactyly and Radial Agenesis Associated with Trisomy 10 Mosaicism

    Directory of Open Access Journals (Sweden)

    Jonathan Lévy

    2013-01-01

    Full Text Available Ectrodactyly or split hand and foot malformations (SHFMs are rare malformations of the limbs, characterized by median clefts of the hands and feet, syndactyly, and aplasia and/or hypoplasia of the phalanges. They represent a clinically and genetically heterogeneous disorder, with both sporadic and familial cases. Most of the genomic rearrangements identified to date in some forms of SHFM are autosomal dominant traits, involving various chromosome regions. Bilateral radial ray defects comprise also a large heterogenous group of disorders, including trisomy 18, Fanconi anemia, and thrombocytopenia-absent-radius syndrome, not commonly associated with ectrodactyly. The present paper describes a case of ectrodactyly associated with bilateral radial ray defects, diagnosed in the first trimester of pregnancy, in a fetus affected by trisomy 10. Only four cases of sporadic and isolated ectrodactyly, diagnosed by ultrasonography between 14 and 22 weeks’ gestation, have been reported. To our knowledge, the present case is the first report of mosaic trisomy 10 associated with SHFM and radial aplasia. Trisomy 10 is a rare lethal chromosomal abnormality, most frequently found in abortion products. Only six liveborn mosaic trisomy 10 infants, with severe malformations, dead in early infancy, have been reported. A severe clinical syndrome can be defined, comprising ear abnormalities, cleft lip/palate, malformations of eyes, heart, and kidneys, and deformity of hands and feet and most often associated with death neonatally or in early infancy.

  15. 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

  16. 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

  17. Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

    International Nuclear Information System (INIS)

    LaFontaine, M.W.R.; Zeller, M.B.; McPhee, G.P.

    2011-01-01

    Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

  18. Defueled channel experiments in ZED-2 in support of ACR-1000 ROP analysis

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W.R.; Zeller, M.B.; McPhee, G.P. [Atomic Energy of Canada Limited (Canada)

    2011-07-01

    Defueled channel experiments were performed in ZED-2 to help resolve discrepancies between calculated flux detector response during refueling in ACR-1000 according the reactor codes RFSP and MCNP. The data produced from these experiments was later used in a separate Regional-Over-Power (ROP) analysis to verify MCNP and RFSP neutron response predictions during refueling. These experiments provided information on thermal flux distributions interior and exterior to a fueled and defueled channel; and on epithermal absolute flux distributions exterior to the same channel. Critical height and moderator temperature data for fueled and defueled channel conditions were also measured. In addition, standard platinum-clad Inconel Self-Powered Detector (SPD) performance data was obtained. The following reactor physics and SPD parameters were measured in these experiments: C Radial flux distribution inside the channel of interest (fueled and defueled), C Radial flux distribution outside the channel of interest (fueled and defueled), C Epithermal radial flux distribution outside the channel of interest (fueled and defueled), and C SPD response parallel to and normal to the channel of interest (fueled and defueled).

  19. Transient well flow in vertically heterogeneous aquifers

    Science.gov (United States)

    Hemker, C. J.

    1999-11-01

    A solution for the general problem of computing well flow in vertically heterogeneous aquifers is found by an integration of both analytical and numerical techniques. The radial component of flow is treated analytically; the drawdown is a continuous function of the distance to the well. The finite-difference technique is used for the vertical flow component only. The aquifer is discretized in the vertical dimension and the heterogeneous aquifer is considered to be a layered (stratified) formation with a finite number of homogeneous sublayers, where each sublayer may have different properties. The transient part of the differential equation is solved with Stehfest's algorithm, a numerical inversion technique of the Laplace transform. The well is of constant discharge and penetrates one or more of the sublayers. The effect of wellbore storage on early drawdown data is taken into account. In this way drawdowns are found for a finite number of sublayers as a continuous function of radial distance to the well and of time since the pumping started. The model is verified by comparing results with published analytical and numerical solutions for well flow in homogeneous and heterogeneous, confined and unconfined aquifers. Instantaneous and delayed drainage of water from above the water table are considered, combined with the effects of partially penetrating and finite-diameter wells. The model is applied to demonstrate that the transient effects of wellbore storage in unconfined aquifers are less pronounced than previous numerical experiments suggest. Other applications of the presented solution technique are given for partially penetrating wells in heterogeneous formations, including a demonstration of the effect of decreasing specific storage values with depth in an otherwise homogeneous aquifer. The presented solution can be a powerful tool for the analysis of drawdown from pumping tests, because hydraulic properties of layered heterogeneous aquifer systems with

  20. 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

  1. Radial head button holing: a cause of irreducible anterior radial head dislocation

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Su-Mi; Chai, Jee Won; You, Ja Yeon; Park, Jina [Seoul National University Seoul Metropolitan Government Boramae Medical Center, Department of Radiology, Seoul (Korea, Republic of); Bae, Kee Jeong [Seoul National University Seoul Metropolitan Government Boramae Medical Center, Department of Orthopedic Surgery, Seoul (Korea, Republic of)

    2016-10-15

    ''Buttonholing'' of the radial head through the anterior joint capsule is a known cause of irreducible anterior radial head dislocation associated with Monteggia injuries in pediatric patients. To the best of our knowledge, no report has described an injury consisting of buttonholing of the radial head through the annular ligament and a simultaneous radial head fracture in an adolescent. In the present case, the radiographic findings were a radial head fracture with anterior dislocation and lack of the anterior fat pad sign. Magnetic resonance imaging (MRI) clearly demonstrated anterior dislocation of the fractured radial head through the torn annular ligament. The anterior joint capsule and proximal portion of the annular ligament were interposed between the radial head and capitellum, preventing closed reduction of the radial head. Familiarity with this condition and imaging findings will aid clinicians to make a proper diagnosis and fast decision to perform an open reduction. (orig.)

  2. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

  3. DDG-1000 Zumwalt Class Destroyer (DDG-1000)

    Science.gov (United States)

    2015-12-01

    Selected Acquisition Report (SAR) RCS: DD-A&T(Q&A)823-197 DDG 1000 Zumwalt Class Destroyer (DDG 1000 ) As of FY 2017 President’s Budget Defense...Acquisition Management Information Retrieval (DAMIR) March 23, 2016 15:17:06 UNCLASSIFIED DDG 1000 December 2015 SAR March 23, 2016 15:17:06...Requirements Document OSD - Office of the Secretary of Defense O&S - Operating and Support PAUC - Program Acquisition Unit Cost DDG 1000 December 2015 SAR

  4. 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

  5. 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

  6. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  7. Radial distribution of the contributions to band broadening of a silica-based semi-preparative monolithic column.

    Science.gov (United States)

    Abia, Jude A; Mriziq, Khaled S; Guiochon, Georges A

    2009-04-01

    Using an on-column local electrochemical microdetector operated in the amperometric mode, band elution profiles were recorded at different radial locations at the exit of a 10 mm id, 100 mm long silica-based monolithic column. HETP plots were then acquired at each of these locations, and all these results were fitted to the Knox equation. This provided a spatial distribution of the values of the eddy diffusion (A), the molecular diffusion (B), and the resistance to the kinetics of mass transfer (C) terms. Results obtained indicate that the wall region yields higher A values and smaller C values than the central core region. Significant radial fluctuations of these contributions to band broadening occur throughout the exit column cross-section. This phenomenon is due to the structural radial heterogeneity of the column.

  8. Impacts on human health from the coal and nuclear fuel cycles and other technologies associated with electric power generation and transmission

    International Nuclear Information System (INIS)

    Radford, E.P.

    1980-01-01

    Major public health impacts of electric power generation and transmission associated with the nuclear fuel cycle and with coal use are evaluated. Only existing technology is evaluated. The only health effects of concern are those leading to definable human disease and injury. Health effects are scaled to a nominal 1000 Megawatt (electric) plant fueled by either option. Comparison of the total health effects to the general public gives: nuclear, 0.03 to 0.05 major health effects per 1000 MWe per year; coal, 0.7 to 3.7 per 1000 MWe per year. Thus for the general public the health risks from the coal cycle are about 50 times greater than for the nuclear cycle. Health effects to workers in the industry are currently quite high. For the nuclear cycle, 4.6 to 5.1 major health impacts per 1000 MWe per year; for coal, 6.5 to 10.9. The two-fold greater risk for the coal cycle is primarily due to high injury rates in coal miners. There is no evidence that electrical transmission contributes any health effects to the general public, except for episodes where broken power lines come in contact with people. For power line workers, the risk is estimated at 0.1 serious injury per 1000 MWe per year

  9. Neutronic feasibility of an LMFBR super long-life core (SLLC)

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Aoki, Katsutada; Arie, Kazuo; Tsuboi, Yasushi

    1988-01-01

    The LMFBR Super Long-Life Core (SLLC) concept has evolved over the last few years as one of the targets of innovative approaches for future FBR cost reduction. An idea for SLLC has been developed wherein the core lifetime is extended up to the plant life of about 30 years by applying the radially and axially multi-zoned core concept (the improved homogeneous core concept). The main purpose of the present study is placed on the evaluation of neutronic feasibility of the 1000 MWe class SLLC concept. The core size of the present SLLC, which is approximately 3 to 4 times as large as those of the current 1000 MWe core design, was determined by the limit of the maximum fast neutron fluence level, which was tentatively assumed to be 5-6x10 23 nvt as the target of the future development of advanced cladding materials. Emphasis is placed on the discussion of neutronic performances of cores with oxide fuels rather than metal or carbide fuels. The present study has shown that proper zoning of the different plutonium enrichment fuels at the initial core makes it possible to achieve small enough reactivity loss during 30-year burnup while satisfying mild variation of the subassembly power distributions using a higher fuel volume fraction of about 50%. Effects of important neutronic parameters on the core performances are also discussed. (orig.)

  10. Minimum throttling feedwater control in VVER-1000 and PWR NPPs

    International Nuclear Information System (INIS)

    Symkin, B.E.; Thaulez, F.

    2004-01-01

    This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

  11. Stability of radial and non-radial pulsation modes of massive ZAMS models

    International Nuclear Information System (INIS)

    Odell, A.P.; Pausenwein, A.; Weiss, W.W.; Hajek, A.

    1987-01-01

    The authors have computed non-adiabatic eigenvalues for radial and non-radial pulsation modes of star models between 80 and 120 M solar with composition of chi=0.70 and Z=0.02. The radial fundamental mode is unstable in models with mass greater than 95 M solar , but the first overtone mode is always stable. The non-radial modes are all stable for all models, but the iota=2 f-mode is the closest to being driven. The non-radial modes are progressively more stable with higher iota and with higher n (for both rho- and g-modes). Thus, their results indicate that radial pulsation limits the upper mass of a star

  12. Development of a VVER-1000 core loading pattern optimization program based on perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2012-01-01

    Highlights: ► We use perturbation theory to find an optimum fuel loading pattern in a VVER-1000. ► We provide a software for in-core fuel management optimization. ► We consider two objectives for our method (perturbation theory). ► We show that perturbation theory method is very fast and accurate for optimization. - Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

  13. Core/shell magnetic mesoporous silica nanoparticles with radially oriented wide mesopores

    Directory of Open Access Journals (Sweden)

    Nikola Ž. Knežević

    2014-06-01

    Full Text Available Core/shell nanoparticles, containing magnetic iron-oxide (maghemite core and mesoporous shell with radial porous structure, were prepared by dispersing magnetite nanoparticles and adding tetraethylorthosilicate to a basic aqueous solution containing structure-templating cetyltrimethylammonium bromide and a pore-swelling mesithylene. The material is characterized by SEM and TEM imaging, nitrogen sorption and powder X-ray diffraction. Distinctive features of the prepared material are its high surface area (959 m2/g, wide average pore diameter (12.4 nm and large pore volume (2.3 cm3/g. The material exhibits radial pore structure and the high angle XRD pattern characteristic for maghemite nanoparticles, which are obtained upon calcination of the magnetite-containing material. The observed properties of the prepared material may render the material applicable in separation, drug delivery, sensing and heterogeneous catalysis.

  14. 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)

  15. 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)

  16. 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)

  17. 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

  18. 17 CFR 229.1000 - (Item 1000) Definitions.

    Science.gov (United States)

    2010-04-01

    ... 17 Commodity and Securities Exchanges 2 2010-04-01 2010-04-01 false (Item 1000) Definitions. 229.1000 Section 229.1000 Commodity and Securities Exchanges SECURITIES AND EXCHANGE COMMISSION STANDARD INSTRUCTIONS FOR FILING FORMS UNDER SECURITIES ACT OF 1933, SECURITIES EXCHANGE ACT OF 1934 AND ENERGY POLICY AND CONSERVATION ACT OF 1975-REGULATION...

  19. 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)

  20. 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

  1. Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Bhowmik, P. K.; Shamim, J. A.; Gairola, A.; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper

  2. 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.

  3. Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

    International Nuclear Information System (INIS)

    Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

    2006-01-01

    The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

  4. AP1000R design robustness against extreme external events - Seismic, flooding, and aircraft crash

    International Nuclear Information System (INIS)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J.

    2012-01-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000 R nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is

  5. Distinct roles of neuroepithelial-like and radial glia-like progenitor cells in cerebellar regeneration.

    Science.gov (United States)

    Kaslin, Jan; Kroehne, Volker; Ganz, Julia; Hans, Stefan; Brand, Michael

    2017-04-15

    Zebrafish can regenerate after brain injury, and the regenerative process is driven by resident stem cells. Stem cells are heterogeneous in the vertebrate brain, but the significance of having heterogeneous stem cells in regeneration is not understood. Limited availability of specific stem cells might impair the regeneration of particular cell lineages. We studied regeneration of the adult zebrafish cerebellum, which contains two major stem and progenitor cell types: ventricular zone and neuroepithelial cells. Using conditional lineage tracing we demonstrate that cerebellar regeneration depends on the availability of specific stem cells. Radial glia-like cells are thought to be the predominant stem cell type in homeostasis and after injury. However, we find that radial glia-like cells play a minor role in adult cerebellar neurogenesis and in recovery after injury. Instead, we find that neuroepithelial cells are the predominant stem cell type supporting cerebellar regeneration after injury. Zebrafish are able to regenerate many, but not all, cell types in the cerebellum, which emphasizes the need to understand the contribution of different adult neural stem and progenitor cell subtypes in the vertebrate central nervous system. © 2017. Published by The Company of Biologists Ltd.

  6. 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)

  7. 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.

  8. 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

  9. Analysis method for the design of a hydrogen mitigation system with passive autocatalytic recombiners in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C-H.; Sung, J-J.; Ha, S-J. [Korea Hydro and Nuclear Power Co. Ltd., Central Research Inst., Daejeon (Korea, Republic of); Yeo, I-S. [KEPCO Engineering and Construction Co. Ltd, Gyeonggi-do (Korea, Republic of)

    2014-07-01

    The importance of hydrogen safety in nuclear power plants has been emphasized especially after the Fukushima accident in Japan. A passive autocatalytic recombiner (PAR) is considered as a viable option for the mitigation of hydrogen risk because of its passive operation for hydrogen removal. This paper presents a licensed hydrogen analysis method of OPR-1000, a 1,000MWe Korea standardized pressurized water reactor with a large dry containment, to determine the capacity and locations of PARs for the design of a hydrogen mitigation system with PAR. Various accident scenarios have been adopted considering important event sequences from a combination of probabilistic methods, deterministic methods and sound engineering judgment. A MAAP 4.0.6+ with a multi-compartment model is used as an analysis tool with conservative hydrogen generation and removal models. The detailed analyses are performed for selected severe accident scenarios including sensitivity analysis with/without operations of various safety systems. The possibility of global flame acceleration (FA) and deflagration-to-detonation transient (DDT) are assessed with sigma (flame acceleration potential) and 7-lambda (DDT potential) criterion. It is concluded that the newly designed hydrogen mitigation system with twenty-four (24) PARs can effectively remove hydrogen in the containment atmosphere and prevent global FA and DDT. (author)

  10. Crosstalk-Managed Heterogeneous Single-Mode 32-Core Fibre

    DEFF Research Database (Denmark)

    Sasaki, Y.; Fukumoto, Ryohei; Takenaga, Katsuhiro

    2016-01-01

    A heterogeneous single-mode 32-core fibre with a cladding diameter of 243 micrometer is designed and fabricated. The highest core count in single-mode multi-core fibres and low worst-case crosstalk of less than -24 dB/1000 km in C-band are achieved simultaneously....

  11. Calculation of economic and financing of NPP and conventional power plant using spreadsheet innovation

    International Nuclear Information System (INIS)

    Moch Djoko Birmano; Imam Bastori

    2008-01-01

    The study for calculating the economic and financing of Nuclear Power Plant (NPP) and conventional power plant using spreadsheet Innovation has been done. As case study, the NPP of PWR type of class 1050 MWe is represented by OPR-1000 (Optimized Power Reactor, 1000 MWe) and the conventional plant of class 600 MWe, is coal power plant (Coal PP). The purpose of the study is to assess the economic and financial feasibility level of OPR-1000 and Coal PP. The study result concludes that economically, OPR-1000 is more feasible compared to Coal PP because its generation cost is cheaper. Whereas financially, OPR-1000 is more beneficial compared to Coal PP because the higher benefit at the end of economic lifetime (NPV) and the higher ratio of benefit and cost (B/C Ratio). For NPP and Coal PP, the higher Discount Rate (%) is not beneficial. NPP is more sensitive to the change of discount rate compared to coal PP, whereas Coal PP is more sensitive to the change of power purchasing price than NPP. (author)

  12. 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)

  13. 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

  14. Quantifying seismic anisotropy induced by small-scale chemical heterogeneities

    Science.gov (United States)

    Alder, C.; Bodin, T.; Ricard, Y.; Capdeville, Y.; Debayle, E.; Montagner, J. P.

    2017-12-01

    Observations of seismic anisotropy are usually used as a proxy for lattice-preferred orientation (LPO) of anisotropic minerals in the Earth's mantle. In this way, seismic anisotropy observed in tomographic models provides important constraints on the geometry of mantle deformation associated with thermal convection and plate tectonics. However, in addition to LPO, small-scale heterogeneities that cannot be resolved by long-period seismic waves may also produce anisotropy. The observed (i.e. apparent) anisotropy is then a combination of an intrinsic and an extrinsic component. Assuming the Earth's mantle exhibits petrological inhomogeneities at all scales, tomographic models built from long-period seismic waves may thus display extrinsic anisotropy. In this paper, we investigate the relation between the amplitude of seismic heterogeneities and the level of induced S-wave radial anisotropy as seen by long-period seismic waves. We generate some simple 1-D and 2-D isotropic models that exhibit a power spectrum of heterogeneities as what is expected for the Earth's mantle, that is, varying as 1/k, with k the wavenumber of these heterogeneities. The 1-D toy models correspond to simple layered media. In the 2-D case, our models depict marble-cake patterns in which an anomaly in shear wave velocity has been advected within convective cells. The long-wavelength equivalents of these models are computed using upscaling relations that link properties of a rapidly varying elastic medium to properties of the effective, that is, apparent, medium as seen by long-period waves. The resulting homogenized media exhibit extrinsic anisotropy and represent what would be observed in tomography. In the 1-D case, we analytically show that the level of anisotropy increases with the square of the amplitude of heterogeneities. This relation is numerically verified for both 1-D and 2-D media. In addition, we predict that 10 per cent of chemical heterogeneities in 2-D marble-cake models can

  15. 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)

  16. 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)

  17. 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

  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. 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)

  20. 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.)

  1. Study, analysis, assess and compare the nuclear engineering systems of nuclear power plant with different reactor types VVER-1000, namely AES-91, AES-92 and AES-2006

    International Nuclear Information System (INIS)

    Le Van Hong; Tran Chi Thanh; Hoang Minh Giang; Le Dai Dien; Nguyen Nhi Dien; Nguyen Minh Tuan

    2015-01-01

    On November 25, 2009, in Hanoi, the National Assembly had been approved the resolution about policy for investment of nuclear power project in Ninh Thuan province which include two sites, each site has two units with power around 1000 MWe. For the nuclear power project at Ninh Thuan 1, Vietnam Government signed the Joint-Governmental Agreement with Russian Government for building the nuclear power plant with reactor type VVER. At present time, the Russian Consultant proposed four reactor technologies can be used for Ninh Thuan 1 project, namely: AES-91, AES-92, AES-2006/V491 and AES-2006/V392M. This report presents the main reactor engineering systems of nuclear power plants with VVER-1000/1200. The results from analysis, comparison and assessment between the designs of AES-91, AES-92 and AES-2006 are also presented. The obtained results show that the type AES-2006 is appropriate selection for Vietnam. (author)

  2. Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

    International Nuclear Information System (INIS)

    Jati Susilo; Jupiter Sitorus Pane

    2016-01-01

    AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

  3. 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%.

  4. Radial optimization of a BWR fuel cell using genetic algorithms

    International Nuclear Information System (INIS)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P.

    2006-01-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U 235 and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix the placement of

  5. Probing Mantle Heterogeneity Across Spatial Scales

    Science.gov (United States)

    Hariharan, A.; Moulik, P.; Lekic, V.

    2017-12-01

    Inferences of mantle heterogeneity in terms of temperature, composition, grain size, melt and crystal structure may vary across local, regional and global scales. Probing these scale-dependent effects require quantitative comparisons and reconciliation of tomographic models that vary in their regional scope, parameterization, regularization and observational constraints. While a range of techniques like radial correlation functions and spherical harmonic analyses have revealed global features like the dominance of long-wavelength variations in mantle heterogeneity, they have limited applicability for specific regions of interest like subduction zones and continental cratons. Moreover, issues like discrepant 1-D reference Earth models and related baseline corrections have impeded the reconciliation of heterogeneity between various regional and global models. We implement a new wavelet-based approach that allows for structure to be filtered simultaneously in both the spectral and spatial domain, allowing us to characterize heterogeneity on a range of scales and in different geographical regions. Our algorithm extends a recent method that expanded lateral variations into the wavelet domain constructed on a cubed sphere. The isolation of reference velocities in the wavelet scaling function facilitates comparisons between models constructed with arbitrary 1-D reference Earth models. The wavelet transformation allows us to quantify the scale-dependent consistency between tomographic models in a region of interest and investigate the fits to data afforded by heterogeneity at various dominant wavelengths. We find substantial and spatially varying differences in the spectrum of heterogeneity between two representative global Vp models constructed using different data and methodologies. Applying the orthonormality of the wavelet expansion, we isolate detailed variations in velocity from models and evaluate additional fits to data afforded by adding such complexities to long

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Adaptation and heterogeneity of Escherichia coli MC1000 growing in complex environments

    DEFF Research Database (Denmark)

    Puentes-Téllez, Pilar; Hansen, Martin Asser; Sørensen, Søren

    2013-01-01

    In a study aiming to assess bacterial evolution in complex growth media, we evaluated the long-term adaptive response of Escherichia coli MC1000 in Luria-Bertani (LB) medium. Seven parallel populations were founded and followed over 150 days in sequential batch cultures under three different oxygen...... conditions (defined environments), and 19 evolved forms were isolated. The emergence of forms with enhanced fitness was evident in competition experiments of all evolved forms versus the ancestral strain. The evolved forms were then subjected to phenotypic and genomic analyses relative to the ancestor...... in galR, a repressor of the galactose operon. Concomitantly, the new forms revealed enhanced growth on galactose as well as galactose-containing disaccharides. This response was likely driven by the LB medium....

  12. Comparison of CPR1000 and AP1000 rod position indication systems

    International Nuclear Information System (INIS)

    Lei Qing

    2009-01-01

    This paper introduces the structure, the function, the digital detection principle of reactor control rod position and monitoring systems in CPR1000 and AP1000, comparing with the characteristics of the system design. The results show that the operation mode and function of AP1000 Rod position indication system are similar to that of CPR1000, but AP1000 rod position system provides higher reliability, and reduces the numbers of containment electrical penetrations and is with better characteristics than that of CPR1000, since it incorporated the redundancy design and data communication. (authors)

  13. 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.

  14. 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

  15. 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

  16. Hot radial pressing: An alternative technique for the manufacturing of plasma-facing components

    International Nuclear Information System (INIS)

    Visca, E.; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A.; Testani, C.

    2005-01-01

    The Hot radial pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock mock-ups were successfully tested to high heat flux thermal fatigue (20 MW/m 2 of absorbed heat flux for 1000 cycles). After these good results the activity is now focused on the developing of a manufacturing process suitable also for the CFC monoblock mock-ups. A FE calculation was performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture. The results obtained by the FE calculation and by the test performed in air simulating a HRP manufacturing process for a CFC monoblock mock-ups is reported in the paper

  17. Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  18. Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

    Kirilova, D.; Belousov, S.; Ilieva, K.

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  19. 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

  20. 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

  1. AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel

  2. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  3. General thermo-elastic solution of radially heterogeneous, spherically isotropic rotating sphere

    Energy Technology Data Exchange (ETDEWEB)

    Bayat, Yahya; EkhteraeiToussi, THamid [Ferdowsi University of Mashhad, Mashhad (Iran, Islamic Republic of)

    2015-06-15

    A thick walled rotating spherical object made of transversely isotropic functionally graded materials (FGMs) with general types of thermo-mechanical boundary conditions is studied. The thermo-mechanical governing equations consisting of decoupled thermal and mechanical equations are represented. The centrifugal body forces of the rotation are considered in the modeling phase. The unsymmetrical thermo-mechanical boundary conditions and rotational body forces are expressed in terms of the Legendre series. The series method is also implemented in the solution of the resulting equations. The solutions are checked with the known literature and FEM based solutions of ABAQUS software. The effects of anisotropy and heterogeneity are studied through the case studies and the results are represented in different figures. The newly developed series form solution is applicable to the rotating FGM spherical transversely isotropic vessels having nonsymmetrical thermo-mechanical boundary condition.

  4. Numerical investigations on axial and radial blade rubs in turbo-machinery

    Science.gov (United States)

    Abdelrhman, Ahmed M.; Tang, Eric Sang Sung; Salman Leong, M.; Al-Qrimli, Haidar F.; Rajamohan, G.

    2017-07-01

    In the recent years, the clearance between the rotor blades and stator/casing had been getting smaller and smaller prior improving the aerodynamic efficiency of the turbomachines as demand in the engineering field. Due to the clearance reduction between the blade tip and the rotor casing and between rotor blades and stator blades, axial and radial blade rubbing could be occurred, especially at high speed resulting into complex nonlinear vibrations. The primary aim of this study is to address the blade axial rubbing phenomenon using numerical analysis of rotor system. A comparison between rubbing caused impacts of axial and radial blade rubbing and rubbing forces are also aims of this study. Tow rotor models (rotor-stator and rotor casing models) has been designed and sketched using SOILDSWORKS software. ANSYS software has been used for the simulation and the numerical analysis. The rubbing conditions were simulated at speed range of 1000rpm, 1500rpm and 2000rpm. Analysis results for axial blade rubbing showed the appearance of blade passing frequency and its multiple frequencies (lx, 2x 3x etc.) and these frequencies will more excited with increasing the rotational speed. Also, it has been observed that when the rotating speed increased, the rubbing force and the harmonics frequencies in x, y and z-direction become higher and severe. The comparison study showed that axial blade rub is more dangerous and would generate a higher vibration impacts and higher blade rubbing force than radial blade rub.

  5. 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.

  6. An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

    International Nuclear Information System (INIS)

    Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

    1985-01-01

    A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

  7. Antiproton compression and radial measurements

    CERN Document Server

    Andresen, G B; Bowe, P D; Bray, C C; Butler, E; Cesar, C L; Chapman, S; Charlton, M; Fajans, J; Fujiwara, M C; Funakoshi, R; Gill, D R; Hangst, J S; Hardy, W N; Hayano, R S; Hayden, M E; Humphries, A J; Hydomako, R; Jenkins, M J; Jorgensen, L V; Kurchaninov, L; Lambo, R; Madsen, N; Nolan, P; Olchanski, K; Olin, A; Page R D; Povilus, A; Pusa, P; Robicheaux, F; Sarid, E; Seif El Nasr, S; Silveira, D M; Storey, J W; Thompson, R I; Van der Werf, D P; Wurtele, J S; Yamazaki, Y

    2008-01-01

    Control of the radial profile of trapped antiproton clouds is critical to trapping antihydrogen. We report detailed measurements of the radial manipulation of antiproton clouds, including areal density compressions by factors as large as ten, achieved by manipulating spatially overlapped electron plasmas. We show detailed measurements of the near-axis antiproton radial profile, and its relation to that of the electron plasma. We also measure the outer radial profile by ejecting antiprotons to the trap wall using an octupole magnet.

  8. 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)

  9. 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)

  10. 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

  11. Difference of reactor core nuclear instrument between AP1000 and CPR1000

    International Nuclear Information System (INIS)

    Zhang Shidong; Zhou Can; Deng Tian

    2014-01-01

    As a typical generation Ⅲ reactor technique, the AP1000 applies many advanced design concepts, simplifies the design, reduces equipment quantities, and thus enhances systematic reliability. The comparison of reactor core measurement instrument differences between AP1000 and CPR1000 from several aspects was involved in the paper. Through analysis and comparison of these differences, passive design concepts and characteristics of AP1000 are familiarized, and conveniences for staffs engaged in CPR1000 to learn and grasp AP1000 technique are provided. It is useful in reactor start up, operation and maintenance. (authors)

  12. 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)

  13. 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

  14. L1000CDS2: LINCS L1000 characteristic direction signatures search engine.

    Science.gov (United States)

    Duan, Qiaonan; Reid, St Patrick; Clark, Neil R; Wang, Zichen; Fernandez, Nicolas F; Rouillard, Andrew D; Readhead, Ben; Tritsch, Sarah R; Hodos, Rachel; Hafner, Marc; Niepel, Mario; Sorger, Peter K; Dudley, Joel T; Bavari, Sina; Panchal, Rekha G; Ma'ayan, Avi

    2016-01-01

    The library of integrated network-based cellular signatures (LINCS) L1000 data set currently comprises of over a million gene expression profiles of chemically perturbed human cell lines. Through unique several intrinsic and extrinsic benchmarking schemes, we demonstrate that processing the L1000 data with the characteristic direction (CD) method significantly improves signal to noise compared with the MODZ method currently used to compute L1000 signatures. The CD processed L1000 signatures are served through a state-of-the-art web-based search engine application called L1000CDS 2 . The L1000CDS 2 search engine provides prioritization of thousands of small-molecule signatures, and their pairwise combinations, predicted to either mimic or reverse an input gene expression signature using two methods. The L1000CDS 2 search engine also predicts drug targets for all the small molecules profiled by the L1000 assay that we processed. Targets are predicted by computing the cosine similarity between the L1000 small-molecule signatures and a large collection of signatures extracted from the gene expression omnibus (GEO) for single-gene perturbations in mammalian cells. We applied L1000CDS 2 to prioritize small molecules that are predicted to reverse expression in 670 disease signatures also extracted from GEO, and prioritized small molecules that can mimic expression of 22 endogenous ligand signatures profiled by the L1000 assay. As a case study, to further demonstrate the utility of L1000CDS 2 , we collected expression signatures from human cells infected with Ebola virus at 30, 60 and 120 min. Querying these signatures with L1000CDS 2 we identified kenpaullone, a GSK3B/CDK2 inhibitor that we show, in subsequent experiments, has a dose-dependent efficacy in inhibiting Ebola infection in vitro without causing cellular toxicity in human cell lines. In summary, the L1000CDS 2 tool can be applied in many biological and biomedical settings, while improving the extraction of

  15. 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.

  16. Stability of radial swirl flows

    International Nuclear Information System (INIS)

    Dou, H S; Khoo, B C

    2012-01-01

    The energy gradient theory is used to examine the stability of radial swirl flows. It is found that the flow of free vortex is always stable, while the introduction of a radial flow will induce the flow to be unstable. It is also shown that the pure radial flow is stable. Thus, there is a flow angle between the pure circumferential flow and the pure radial flow at which the flow is most unstable. It is demonstrated that the magnitude of this flow angle is related to the Re number based on the radial flow rate, and it is near the pure circumferential flow. The result obtained in this study is useful for the design of vaneless diffusers of centrifugal compressors and pumps as well as other industrial devices.

  17. 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)

  18. 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

  19. 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)

  20. Scalable synthesis and post-modification of a mesoporous metal-organic framework called NU-1000.

    Science.gov (United States)

    Wang, Timothy C; Vermeulen, Nicolaas A; Kim, In Soo; Martinson, Alex B F; Stoddart, J Fraser; Hupp, Joseph T; Farha, Omar K

    2016-01-01

    The synthesis of NU-1000, a highly robust mesoporous (containing pores >2 nm) metal-organic framework (MOF), can be conducted efficiently on a multigram scale from inexpensive starting materials. Tetrabromopyrene and (4-(ethoxycarbonyl)phenyl)boronic acid can easily be coupled to prepare the requisite organic strut with four metal-binding sites in the form of four carboxylic acids, while zirconyl chloride octahydrate is used as a precursor for the well-defined metal oxide clusters. NU-1000 has been reported as an excellent candidate for the separation of gases, and it is a versatile scaffold for heterogeneous catalysis. In particular, it is ideal for the catalytic deactivation of nerve agents, and it shows great promise as a new generic platform for a wide range of applications. Multiple post-synthetic modification protocols have been developed using NU-1000 as the parent material, making it a potentially useful scaffold for several catalytic applications. The procedure for the preparation of NU-1000 can be scaled up reliably, and it is suitable for the production of 50 g of the tetracarboxylic acid containing organic linker and 200 mg-2.5 g of NU-1000. The entire synthesis is performed without purification by column chromatography and can be completed within 10 d.

  1. 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

  2. Cellular modelling of secondary radial growth in conifer trees: application to Pinus radiata (D. Don).

    Science.gov (United States)

    Forest, Loïc; Demongeot, Jacques; Demongeota, Jacques

    2006-05-01

    The radial growth of conifer trees proceeds from the dynamics of a merismatic tissue called vascular cambium or cambium. Cambium is a thin layer of active proliferating cells. The purpose of this paper was to model the main characteristics of cambial activity and its consecutive radial growth. Cell growth is under the control of the auxin hormone indole-3-acetic. The model is composed of a discrete part, which accounts for cellular proliferation, and a continuous part involving the transport of auxin. Cambium is modeled in a two-dimensional cross-section by a cellular automaton that describes the set of all its constitutive cells. Proliferation is defined as growth and division of cambial cells under neighbouring constraints, which can eliminate some cells from the cambium. The cell-growth rate is determined from auxin concentration, calculated with the continuous model. We studied the integration of each elementary cambial cell activity into the global coherent movement of macroscopic morphogenesis. Cases of normal and abnormal growth of Pinus radiata (D. Don) are modelled. Abnormal growth includes deformed trees where gravity influences auxin transport, producing heterogeneous radial growth. Cross-sectional microscopic views are also provided to validate the model's hypothesis and results.

  3. [Comparison of chemical quality characteristics between radial striations and non-radial striations in tuberous root of Rehmannia glutinosa].

    Science.gov (United States)

    Xie, Cai-Xia; Zhang, Miao; Li, Ya-Jing; Geng, Xiao-Tong; Wang, Feng-Qing; Zhang, Zhong-Yi

    2017-11-01

    An HPLC method was established to determine the contents of catalpol, acteoside, rehmaionoside A, rehmaionoside D, leonuride in three part of Rehmanni glutinosa in Beijing No.1 variety R. glutinosa during the growth period, This method, in combination with its HPLC fingerprint was used to evaluate its overall quality characteristics.The results showed that:① the content of main components of R. glutinosa varied in different growth stages ;② there was a great difference of the content of main components between theradial striations and the non-radial striations; ③ the two sections almost have the same content distribution of catalpol, acteoside and rehmaionoside D; ④the content of rehmaionoside A in non-radial striations was higher than that in radial striations,while the content of leonuride in radial striations was higher than that in non-radial striations.; ⑤the HPLC fingerprint of radial striations, non-radial striations and whole root tuber were basically identical, except for the big difference in the content of chemical components. The result of clustering displayed that the radial striations, non-radial striations, and whole root were divided into two groups. In conclusion, there was a significant difference in the quality characteristics of radial striations and non-radial striations of R. glutinosa. This research provides a reference for quality evaluation and geoherbalism of R. glutinosa. Copyright© by the Chinese Pharmaceutical Association.

  4. 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

  5. Experimental feasibility study of radial injection cooling of three-pad radial air foil bearings

    Science.gov (United States)

    Shrestha, Suman K.

    Air foil bearings use ambient air as a lubricant allowing environment-friendly operation. When they are designed, installed, and operated properly, air foil bearings are very cost effective and reliable solution to oil-free turbomachinery. Because air is used as a lubricant, there are no mechanical contacts between the rotor and bearings and when the rotor is lifted off the bearing, near frictionless quiet operation is possible. However, due to the high speed operation, thermal management is one of the very important design factors to consider. Most widely accepted practice of the cooling method is axial cooling, which uses cooling air passing through heat exchange channels formed underneath the bearing pad. Advantage is no hardware modification to implement the axial cooling because elastic foundation structure of foil bearing serves as a heat exchange channels. Disadvantage is axial temperature gradient on the journal shaft and bearing. This work presents the experimental feasibility study of alternative cooling method using radial injection of cooling air directly on the rotor shaft. The injection speeds, number of nozzles, location of nozzles, total air flow rate are important factors determining the effectiveness of the radial injection cooling method. Effectiveness of the radial injection cooling was compared with traditional axial cooling method. A previously constructed test rig was modified to accommodate a new motor with higher torque and radial injection cooling. The radial injection cooling utilizes the direct air injection to the inlet region of air film from three locations at 120° from one another with each location having three axially separated holes. In axial cooling, a certain axial pressure gradient is applied across the bearing to induce axial cooling air through bump foil channels. For the comparison of the two methods, the same amount of cooling air flow rate was used for both axial cooling and radial injection. Cooling air flow rate was

  6. ACT-1000. Group activation cross-section library for WWER-1000 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zolotarev, K I; Pashchenko, A B [National Research Centre - A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-10-01

    The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10{sup -9}-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)

  7. Physical mechanism determining the radial electric field and its radial structure in a toroidal plasma

    International Nuclear Information System (INIS)

    Ida, Katsumi; Miura, Yukitoshi; Itoh, Sanae

    1994-10-01

    Radial structures of plasma rotation and radial electric field are experimentally studied in tokamak, heliotron/torsatron and stellarator devices. The perpendicular and parallel viscosities are measured. The parallel viscosity, which is dominant in determining the toroidal velocity in heliotron/torsatron and stellarator devices, is found to be neoclassical. On the other hand, the perpendicular viscosity, which is dominant in dictating the toroidal rotation in tokamaks, is anomalous. Even without external momentum input, both a plasma rotation and a radial electric field exist in tokamaks and heliotrons/torsatrons. The observed profiles of the radial electric field do not agree with the theoretical prediction based on neoclassical transport. This is mainly due to the existence of anomalous perpendicular viscosity. The shear of the radial electric field improves particle and heat transport both in bulk and edge plasma regimes of tokamaks. (author) 95 refs

  8. An advanced method of heterogeneous reactor theory

    International Nuclear Information System (INIS)

    Kochurov, B.P.

    1994-08-01

    Recent approaches to heterogeneous reactor theory for numerical applications were presented in the course of 8 lectures given in JAERI. The limitations of initial theory known after the First Conference on Peacefull Uses of Atomic Energy held in Geneva in 1955 as Galanine-Feinberg heterogeneous theory:-matrix from of equations, -lack of consistent theory for heterogeneous parameters for reactor cell, -were overcome by a transformation of heterogeneous reactor equations to a difference form and by a development of a consistent theory for the characteristics of a reactor cell based on detailed space-energy calculations. General few group (G-number of groups) heterogeneous reactor equations in dipole approximation are formulated with the extension of two-dimensional problem to three-dimensions by finite Furie expansion of axial dependence of neutron fluxes. A transformation of initial matrix reactor equations to a difference form is presented. The methods for calculation of heterogeneous reactor cell characteristics giving the relation between vector-flux and vector-current on a cell boundary are based on a set of detailed space-energy neutron flux distribution calculations with zero current across cell boundary and G calculations with linearly independent currents across the cell boundary. The equations for reaction rate matrices are formulated. Specific methods were developed for description of neutron migration in axial and radial directions. The methods for resonance level's approach for numerous high-energy resonances. On the basis of these approaches the theory, methods and computer codes were developed for 3D space-time react or problems including simulation of slow processes with fuel burn-up, control rod movements, Xe poisoning and fast transients depending on prompt and delayed neutrons. As a result reactors with several thousands of channels having non-uniform axial structure can be feasibly treated. (author)

  9. Radial retinotomy in the macula.

    Science.gov (United States)

    Bovino, J A; Marcus, D F

    1984-01-01

    Radial retinotomy is an operative procedure usually performed in the peripheral or equatorial retina. To facilitate retinal attachment, the authors used intraocular scissors to perform radial retinotomy in the macula of two patients during vitrectomy surgery. In the first patient, a retinal detachment complicated by periretinal proliferation and macula hole formation was successfully reoperated with the aid of three radial cuts in the retina at the edges of the macular hole. In the second patient, an intraoperative retinal tear in the macula during diabetic vitrectomy was also successfully repaired with the aid of radial retinotomy. In both patients, retinotomy in the macula was required because epiretinal membranes, which could not be easily delaminated, were hindering retinal reattachment.

  10. Radial head dislocation during proximal radial shaft osteotomy.

    Science.gov (United States)

    Hazel, Antony; Bindra, Randy R

    2014-03-01

    The following case report describes a 48-year-old female patient with a longstanding both-bone forearm malunion, who underwent osteotomies of both the radius and ulna to improve symptoms of pain and lack of rotation at the wrist. The osteotomies were templated preoperatively. During surgery, after performing the planned radial shaft osteotomy, the authors recognized that the radial head was subluxated. The osteotomy was then revised from an opening wedge to a closing wedge with improvement of alignment and rotation. The case report discusses the details of the operation, as well as ways in which to avoid similar shortcomings in the future. Copyright © 2014 American Society for Surgery of the Hand. Published by Elsevier Inc. All rights reserved.

  11. Spontaneous mutations in the flhD operon generate motility heterogeneity in Escherichia coli biofilm.

    Science.gov (United States)

    Horne, Shelley M; Sayler, Joseph; Scarberry, Nicholas; Schroeder, Meredith; Lynnes, Ty; Prüß, Birgit M

    2016-11-08

    Heterogeneity and niche adaptation in bacterial biofilm involve changes to the genetic makeup of the bacteria and gene expression control. We hypothesized that i) spontaneous mutations in the flhD operon can either increase or decrease motility and that ii) the resulting motility heterogeneity in the biofilm might lead to a long-term increase in biofilm biomass. We allowed the highly motile E. coli K-12 strain MC1000 to form seven- and fourteen-day old biofilm, from which we recovered reduced motility isolates at a substantially greater frequency (5.4 %) than from a similar experiment with planktonic bacteria (0.1 %). Biofilms formed exclusively by MC1000 degraded after 2 weeks. In contrast, biofilms initiated with a 1:1 ratio of MC1000 and its isogenic flhD::kn mutant remained intact at 4 weeks and the two strains remained in equilibrium for at least two weeks. These data imply that an 'optimal' biofilm may contain a mixture of motile and non-motile bacteria. Twenty-eight of the non-motile MC1000 isolates contained an IS1 element in proximity to the translational start of FlhD or within the open reading frames for FlhD or FlhC. Two isolates had an IS2 and one isolate had an IS5 in the open reading frame for FlhD. An additional three isolates contained deletions that included the RNA polymerase binding site, five isolates contained point mutations and small deletions in the open reading frame for FlhC. The locations of all these mutations are consistent with the lack of motility and further downstream within the flhD operon than previously published IS elements that increased motility. We believe that the location of the mutation within the flhD operon determines whether the effect on motility is positive or negative. To test the second part of our hypothesis where motility heterogeneity in a biofilm may lead to a long-term increase in biofilm biomass, we quantified biofilm biomass by MC1000, MC1000 flhD::kn, and mixtures of the two strains at ratios of 1:1, 10

  12. Influence of chemical heterogeneity of solid solutions on brittleness in chromium steels

    International Nuclear Information System (INIS)

    Madyanov, S.A.; Sedov, V.K.; Apaev, B.A.

    1985-01-01

    The role of chemical heterogeneity of solid solutions in formation of mechanical properties of Kh09, Kh15, Kh20, Kh19N2G5T chromium steels has been investigated. It is established that besides the known regioA of chemical heterogeneity in the vicinity of 475 deg C exists a high-temperature region (1000-1050 deg C), where maximum heteroge=- neity of chromium distribution in solid solution, is observed. Both types of chemical heterogeneity cause essential hardening of alloys, which becomes apparent in abrupt change of capability to microplastic deformation The mechanism of occurrence of the given temper brittleness consists in carbon diffusion into microvolunes enriched in carbide-forming elements

  13. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  14. Radial cracks and fracture mechanism of radially oriented ring 2:17 type SmCo magnets

    International Nuclear Information System (INIS)

    Tian Jianjun; Pan Dean; Zhou Hao; Yin Fuzheng; Tao Siwu; Zhang Shengen; Qu Xuanhui

    2009-01-01

    Radially oriented ring 2:17 type SmCo magnets have different microstructure in the radial direction (easy magnetization) and axial direction (hard magnetization). The structure of the cross-section in radial direction is close-packed atomic plane, which shows cellular microstructure. The microstructure of the cross-section in axial direction consists of a mixture of rhombic microstructure and parallel lamella phases. So the magnets have obvious anisotropy of thermal expansion in different directions. The difference of the thermal expansion coefficients reaches the maximum value at 830-860 deg. C, which leads to radial cracks during quenching. The magnets have high brittlement because there are fewer slip systems in crystal structure. The fracture is brittle cleavage fracture.

  15. Radial anisotropy of Northeast Asia inferred from Bayesian inversions of ambient noise data

    Science.gov (United States)

    Lee, S. J.; Kim, S.; Rhie, J.

    2017-12-01

    The eastern margin of the Eurasia plate exhibits complex tectonic settings due to interactions with the subducting Pacific and Philippine Sea plates and the colliding India plate. Distributed extensional basins and intraplate volcanoes, and their heterogeneous features in the region are not easily explained with a simple mechanism. Observations of radial anisotropy in the entire lithosphere and the part of the asthenosphere provide the most effective evidence for the deformation of the lithosphere and the associated variation of the lithosphere-asthenosphere boundary (LAB). To infer anisotropic structures of crustal and upper-mantle in this region, radial anisotropy is measured using ambient noise data. In a continuation of previous Rayleigh wave tomography study in Northeast Asia, we conduct Love wave tomography to determine radial anisotropy using the Bayesian inversion techniques. Continuous seismic noise recordings of 237 broad-band seismic stations are used and more than 55,000 group and phase velocities of fundamental mode are measured for periods of 5-60 s. Total 8 different types of dispersion maps of Love wave from this study (period 10-60 s), Rayleigh wave from previous tomographic study (Kim et al., 2016; period 8-70 s) and longer period data (period 70-200 s) from a global model (Ekstrom, 2011) are jointly inverted using a hierarchical and transdimensional Bayesian technique. For each grid-node, boundary depths, velocities and anisotropy parameters of layers are sampled simultaneously on the assumption of the layered half-space model. The constructed 3-D radial anisotropy model provides much more details about the crust and upper mantle anisotropic structures, and about the complex undulation of the LAB.

  16. XRAY applied program package for calculation of electron-photon fields in the energy range of 1-1000 keV

    International Nuclear Information System (INIS)

    Lappa, A.V.; Khadyeva, Z.M.; Burmistrov, D.S.; Vasil'ev, O.N.

    1990-01-01

    The package of applied XRAY programs is intended for calculating the linear and fluctuation characteristics of photon and electron radiation fields in heterogeneous medium within 1-1000 keV energy range. The XRAY program package consists of moduli written in FORTRAN-IV and data files. 9 refs

  17. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  18. 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

  19. 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)

  20. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  1. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  2. 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)

  3. 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)

  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. Perceived radial translation during centrifugation

    NARCIS (Netherlands)

    Bos, J.E.; Correia Grácio, B.J.

    2015-01-01

    BACKGROUND: Linear acceleration generally gives rise to translation perception. Centripetal acceleration during centrifugation, however, has never been reported giving rise to a radial, inward translation perception. OBJECTIVE: To study whether centrifugation can induce a radial translation

  6. Operational benchmark for VVER-1000, unit 6, Kozloduy NPP

    International Nuclear Information System (INIS)

    Apostolov, T.; Petrov, B.

    1999-01-01

    Benchmark calculations have been carried out using the 3D nodal code TRAPEZ. Global neutron-physics characteristics of the VVER-1000 core, Kozloduy NPP Unit 6, have been determined taking into account the real loading patterns and operational history of the first three cycles. The code TRLOAD has been used to perform the fuel reloading between any two cycles. The reactor and components descriptions as well as material compositions are given. The results presented include the critical boric acid concentration, the radial power distribution, the axial power distribution for the maximum overload assembly, and the burnup distribution at three different moments during each cycle. Calculated values have been compared with measured data. It is shown that the results obtained by the TRAPEZ code are in good agreement with the experimental data. The information presented could serve as a test case for validation of code packages designed for analyzing the steady-state operation of VVERs. (author)

  7. Evidence for a Significant Level of Extrinsic Anisotropy Due to Heterogeneities in the Mantle.

    Science.gov (United States)

    Alder, C.; Bodin, T.; Ricard, Y. R.; Capdeville, Y.; Debayle, E.; Montagner, J. P.

    2017-12-01

    Observations of seismic anisotropy are used as a proxy for lattice-preferred orientation (LPO) of anisotropic minerals in the Earth's mantle. In this way, it provides important constraints on the geometry of mantle deformation. However, in addition to LPO, small-scale heterogeneities that cannot be resolved by long-period seismic waves may also produce anisotropy. The observed (i.e. apparent) anisotropy is then a combination of an intrinsic and an extrinsic component. Assuming the Earth's mantle exhibits petrological inhomogeneities at all scales, tomographic models built from long-period seismic waves may thus display extrinsic anisotropy. Here, we investigate the relation between the amplitude of seismic heterogeneities and the level of induced S-wave radial anisotropy as seen by long-period seismic waves. We generate some simple 1D and 2D isotropic models that exhibit a power spectrum of heterogeneities as what is expected for the Earth's mantle, i.e. varying as 1/k, with k the wavenumber of these heterogeneities. The 1D toy models correspond to simple layered media. In the 2D case, our models depict marble-cake patterns in which an anomaly in S-wave velocity has been advected within convective cells. The long-wavelength equivalents of these models are computed using upscaling relations that link properties of a rapidly varying elastic medium to properties of the effective, i.e. apparent, medium as seen by long-period waves. The resulting homogenized media exhibit extrinsic anisotropy and represent what would be observed in tomography. In the 1D case, we analytically show that the level of anisotropy increases with the square of the amplitude of heterogeneities. This relation is numerically verified for both 1D and 2D media. In addition, we predict that 10 % of chemical heterogeneities in 2D marble-cake models can induce more than 3.9 % of extrinsic radial S-wave anisotropy. We thus predict that a non-negligible part of the observed anisotropy in tomographic

  8. Self-consistent radial sheath

    International Nuclear Information System (INIS)

    Hazeltine, R.D.

    1988-12-01

    The boundary layer arising in the radial vicinity of a tokamak limiter is examined, with special reference to the TEXT tokamak. It is shown that sheath structure depends upon the self-consistent effects of ion guiding-center orbit modification, as well as the radial variation of E /times/ B-induced toroidal rotation. Reasonable agreement with experiment is obtained from an idealized model which, however simplified, preserves such self-consistent effects. It is argued that the radial sheath, which occurs whenever confining magnetic field-lines lie in the plasma boundary surface, is an object of some intrinsic interest. It differs from the more familiar axial sheath because magnetized charges respond very differently to parallel and perpendicular electric fields. 11 refs., 1 fig

  9. 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

  10. Radial pseudoaneurysm following diagnostic coronary angiography

    Directory of Open Access Journals (Sweden)

    Shankar Laudari

    2015-06-01

    Full Text Available The radial artery access has gained popularity as a method of diagnostic coronary catheterization compared to femoral artery puncture in terms of vascular complications and early ambulation. However, very rare complication like radial artery pseudoaneurysm may occur following cardiac catheterization which may give rise to serious consequences. Here, we report a patient with radial pseudoaneurysm following diagnostic coronary angiography. Adequate and correct methodology of compression of radial artery following puncture for maintaining hemostasis is the key to prevention.DOI: http://dx.doi.org/10.3126/jcmsn.v10i3.12776 Journal of College of Medical Sciences-Nepal, 2014, Vol-10, No-3, 48-50

  11. Application of heterogeneous method for the interpretation of exponential experiments

    International Nuclear Information System (INIS)

    Birkhoff, G.; Bondar, L.

    1977-01-01

    The present paper gives a brief review of a work which was executed mainly during 1967 and 1968 in the field of the application of heterogeneous methods for the interpretation of exponential experiments with ORGEL type lattices (lattices of natural uranium cluster elements with organic coolants moderated by heavy water). In the frame of this work a heterogeneous computer program, in (r,γ) geometry was written which is based on the NORDHEIM method using a uniform moderator, three energy groups and monopol and dipol sources. This code is especially adapted for regular square lattices in a cylindrical tank. Full use of lattice symmetry was made for reducing the numerical job of the theory. A further reduction was obtained by introducing a group averaged extrapolation distance at the external boundary. Channel parameters were evaluated by the PINOCCHIO code. Comparisons of calculated and measured thermal neutron flux showed good agreement. Equivalence of heterogeneous and homogeneous theory was found in cases of lattices comprising a minimum of 32, 24 and 16 fuel elements for respectively under-, well-, and over-moderated lattices. Heterogeneous calculations of high leakage lattices suffered the lack of good methods for the computation of axial and radial streaming parameters. Interpretation of buckling measurements in the subcritical facility EXPO requires already more accurate evaluation of the streaming effects than we made. The potential of heterogeneous theory in the field of exponential experiments is thought to be limited by the precision by which the streaming parameters can be calculated

  12. Finite volume approximation of the three-dimensional flow equation in axisymmetric, heterogeneous porous media based on local analytical solution

    KAUST Repository

    Salama, Amgad

    2013-09-01

    In this work the problem of flow in three-dimensional, axisymmetric, heterogeneous porous medium domain is investigated numerically. For this system, it is natural to use cylindrical coordinate system, which is useful in describing phenomena that have some rotational symmetry about the longitudinal axis. This can happen in porous media, for example, in the vicinity of production/injection wells. The basic feature of this system is the fact that the flux component (volume flow rate per unit area) in the radial direction is changing because of the continuous change of the area. In this case, variables change rapidly closer to the axis of symmetry and this requires the mesh to be denser. In this work, we generalize a methodology that allows coarser mesh to be used and yet yields accurate results. This method is based on constructing local analytical solution in each cell in the radial direction and moves the derivatives in the other directions to the source term. A new expression for the harmonic mean of the hydraulic conductivity in the radial direction is developed. Apparently, this approach conforms to the analytical solution for uni-directional flows in radial direction in homogeneous porous media. For the case when the porous medium is heterogeneous or the boundary conditions is more complex, comparing with the mesh-independent solution, this approach requires only coarser mesh to arrive at this solution while the traditional methods require more denser mesh. Comparisons for different hydraulic conductivity scenarios and boundary conditions have also been introduced. © 2013 Elsevier B.V.

  13. Endoscopic versus open radial artery harvest and mammario-radial versus aorto-radial grafting in patients undergoing coronary artery bypass surgery

    DEFF Research Database (Denmark)

    Carranza, Christian L; Ballegaard, Martin; Werner, Mads U

    2014-01-01

    the postoperative complications will be registered, and we will evaluate muscular function, scar appearance, vascular supply to the hand, and the graft patency including the patency of the central radial artery anastomosis. A patency evaluation by multi-slice computer tomography will be done at one year...... to aorto-radial revascularisation techniques but this objective is exploratory. TRIAL REGISTRATION: ClinicalTrials.gov identifier: NCT01848886.Danish Ethics committee number: H-3-2012-116.Danish Data Protection Agency: 2007-58-0015/jr.n:30-0838....

  14. 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)

  15. 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.

  16. Numerical simulation of liquid-metal-flows in radial-toroidal-radial bends

    International Nuclear Information System (INIS)

    Molokov, S.; Buehler, L.

    1993-09-01

    Magnetohydrodynamic flows in a U-bend and right-angle bend are considered with reference to the radial-toroidal-radial concept of a self-cooled liquid-metal blanket. The ducts composing bends have rectangular cross-section. The applied magnetic field is aligned with the toroidal duct and perpendicular to the radial ones. At high Hartmann number the flow region is divided into cores and boundary layers of different types. The magnetohydrodynamic equations are reduced to a system of partial differential equations governing wall electric potentials and the core pressure. The system is solved numerically by two different methods. The first method is iterative with iteration between wall potential and the core pressure. The second method is a general one for the solution of the core flow equations in curvilinear coordinates generated by channel geometry and magnetic field orientation. Results obtained are in good agreement. They show, that the 3D-pressure drop of MHD flows in a U-bend is not a critical issue for blanket applications. (orig./HP) [de

  17. ANALISIS SEBARAN RADIONUKLIDA PADA KONDISI NORMAL UNTUK REAKTOR AEC 1000 MW

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2015-03-01

    Full Text Available Telah dilakukan analisis sebaran radionuklida pada reaktor daya Atomic Energy Agency (AEC 3568 MWTh, setara dengan 1000 Mwe untuk kondisi operasi normal. Analisis dilakukan untuk dua reaktor yang terpisah sejauh 500 m dan sudut 90o satu dengan yang lain. Langkah awal dalam melakukan analisis adalah menentukan suku sumber reaktor menggunakan program komputer ORIGEN2 dan EMERALD NORMAL. ORIGEN2 digunakan untuk menentukan inventori radionuklida yang terdapat di reaktor. Selanjutnya dengan dengan menggunakan program EMERALD NORMAL dihitung suku sumber yang sampai ke cerobong reaktor. Untuk menganalisis dosis yang diterima penduduk dilakukan dengan menggunakan program PC-CREAM. Perhitungan dilakukan untuk satu dan dua PLTN di calon tapak PLTN. Hasil yang diperoleh adalah sebaran radionuklida terbesar untuk satu PLTN pada jarak 1 km dan kearah zona 9 (191,25o dan untuk dua PLTN pada jarak 1 km dan kearah zona 10 (213,75o. Radionuklida yang sampai ke penduduk melalui dua alur yaitu alur makanan dan hirupan. Untuk alur makanan berasal dari radionuklida I-131, dan terbesar melalui alur produk susu sebesar 53,40 % untuk satu maupun dua PLTN . Untuk alur hirupan ranionuklida pemberi kontribusi paparan terbesar berasal dari Kr-85m sebesar 53,80 %. Dosis total terbesar yang diterima penduduk terdapat pada jarak 1 Km untuk bayi yaitu sebesar 4,10 μSi dan 11,26 μSi untuk satu dan dua PLTN. Hasil ini sangat kecil dibandingkan dengan batas dosis yang diijinkan oleh badan pengawas (BAPETEN untuk penduduk yaitu sebesar 1 mSi. Kata Kunci : Reaktor daya, komputer code, radionuklida, alur makanan, hirupan   Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC 3568 MWth Power Reactor, equal to the 1000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90o one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL

  18. Dedicated radial ventriculography pigtail catheter

    Energy Technology Data Exchange (ETDEWEB)

    Vidovich, Mladen I., E-mail: miv@uic.edu

    2013-05-15

    A new dedicated cardiac ventriculography catheter was specifically designed for radial and upper arm arterial access approach. Two catheter configurations have been developed to facilitate retrograde crossing of the aortic valve and to conform to various subclavian, ascending aortic and left ventricular anatomies. The “short” dedicated radial ventriculography catheter is suited for horizontal ascending aortas, obese body habitus, short stature and small ventricular cavities. The “long” dedicated radial ventriculography catheter is suited for vertical ascending aortas, thin body habitus, tall stature and larger ventricular cavities. This new design allows for improved performance, faster and simpler insertion in the left ventricle which can reduce procedure time, radiation exposure and propensity for radial artery spasm due to excessive catheter manipulation. Two different catheter configurations allow for optimal catheter selection in a broad range of patient anatomies. The catheter is exceptionally stable during contrast power injection and provides equivalent cavity opacification to traditional femoral ventriculography catheter designs.

  19. Radial wave crystals: radially periodic structures from anisotropic metamaterials for engineering acoustic or electromagnetic waves.

    Science.gov (United States)

    Torrent, Daniel; Sánchez-Dehesa, José

    2009-08-07

    We demonstrate that metamaterials with anisotropic properties can be used to develop a new class of periodic structures that has been named radial wave crystals. They can be sonic or photonic, and wave propagation along the radial directions is obtained through Bloch states like in usual sonic or photonic crystals. The band structure of the proposed structures can be tailored in a large amount to get exciting novel wave phenomena. For example, it is shown that acoustical cavities based on radial sonic crystals can be employed as passive devices for beam forming or dynamically orientated antennas for sound localization.

  20. Aneurisma idiopático de artéria radial: relato de caso Idiopathic radial artery aneurysm: case report

    Directory of Open Access Journals (Sweden)

    Luiz Ernani Meira Jr.

    2011-12-01

    Full Text Available Os aneurismas da artéria radial são extremamente raros. Em sua maioria, consistem de pseudoaneurismas pós-traumáticos. Os aneurismas da artéria radial verdadeiros podem ser idiopáticos, congênitos, pós-estenóticos ou associados a patologias, tais como vasculites e doenças do tecido conjuntivo. Foi relatado um caso de aneurisma idiopático de artéria radial em uma criança de três anos, que, após completa investigação diagnóstica complementar, foi submetida à ressecção cirúrgica.Radial artery aneurysms are extremely rare. Post-traumatic pseudoaneurysms are the vast majority. True radial artery aneurysms can be idiopathic, congenital, poststenotic, or associated with some pathologies, such as vasculitis and conjunctive tissue diseases. We report a case of an idiopathic aneurysm of the radial artery in a three-year-old child who was submitted to surgical resection after a complete diagnostic approach.

  1. Ulnar nerve entrapment complicating radial head excision

    Directory of Open Access Journals (Sweden)

    Kevin Parfait Bienvenu Bouhelo-Pam

    Full Text Available Introduction: Several mechanisms are involved in ischemia or mechanical compression of ulnar nerve at the elbow. Presentation of case: We hereby present the case of a road accident victim, who received a radial head excision for an isolated fracture of the radial head and complicated by onset of cubital tunnel syndrome. This outcome could be the consequence of an iatrogenic valgus of the elbow due to excision of the radial head. Hitherto the surgical treatment of choice it is gradually been abandoned due to development of radial head implant arthroplasty. However, this management option is still being performed in some rural centers with low resources. Discussion: The radial head plays an important role in the stability of the elbow and his iatrogenic deformity can be complicated by cubital tunnel syndrome. Conclusion: An ulnar nerve release was performed with favorable outcome. Keywords: Cubital tunnel syndrome, Peripheral nerve palsy, Radial head excision, Elbow valgus

  2. Insights into the radial water jet drilling technology – Application in a quarry

    Directory of Open Access Journals (Sweden)

    Thomas Reinsch

    2018-04-01

    Full Text Available In this context, we applied the radial water jet drilling (RJD technology to drill five horizontal holes into a quarry wall of the Gildehaus quarry close to Bad Bentheim, Germany. For testing the state-of-the-art jetting technology, a jetting experiment was performed to investigate the influence of geological heterogeneity on the jetting performance and the hole geometry, the influence of nozzle geometry and jetting pressure on the rate of penetration, and the possibility of localising the jetting nozzle utilizing acoustic activity. It is observed that the jetted holes can intersect fractures under varying angles, and the jetted holes do not follow a straight path when jetting at ambient surface condition. Cuttings from the jetting process retrieved from the holes can be used to estimate the reservoir rock permeability. Within the quarry, we did not observe a change in the rate of penetration due to jetting pressure variations. Acoustic monitoring was partially successful in estimating the nozzle location. Although the experiments were performed at ambient surface conditions, the results can give recommendations for a downhole application in deep wells. Keywords: Acoustic monitoring, Drilling performance, Trajectory, Permeability, Rock properties, Radial water jet drilling (RJD

  3. Role of a Modulator in the Synthesis of Phase-Pure NU-1000.

    Science.gov (United States)

    Webber, Thomas E; Liu, Wei-Guang; Desai, Sai Puneet; Lu, Connie C; Truhlar, Donald G; Penn, R Lee

    2017-11-15

    NU-1000 is a robust, mesoporous metal-organic framework (MOF) with hexazirconium nodes ([Zr 6 O 16 H 16 ] 8+ , referred to as oxo-Zr 6 nodes) that can be synthesized by combining a solution of ZrOCl 2 ·8H 2 O and a benzoic acid modulator in N,N-dimethylformamide with a solution of linker (1,3,6,8-tetrakis(p-benzoic acid)pyrene, referred to as H 4 TBAPy) and by aging at an elevated temperature. Typically, the resulting crystals are primarily composed of NU-1000 domains that crystallize with a more dense phase that shares structural similarity with NU-901, which is an MOF composed of the same linker molecules and nodes. Density differences between the two polymorphs arise from the differences in the node orientation: in NU-1000, the oxo-Zr 6 nodes rotate 120° from node to node, whereas in NU-901, all nodes are aligned in parallel. Considering this structural difference leads to the hypothesis that changing the modulator from benzoic acid to a larger and more rigid biphenyl-4-carboxylic acid might lead to a stronger steric interaction between the modulator coordinating on the oxo-Zr 6 node and misaligned nodes or linkers in the large pore and inhibit the growth of the more dense NU-901-like material, resulting in phase-pure NU-1000. Side-by-side reactions comparing the products of synthesis using benzoic acid or biphenyl-4-carboxylic acid as a modulator produce structurally heterogeneous crystals and phase-pure NU-1000 crystals. It can be concluded that the larger and more rigid biphenyl-4-carboxylate inhibits the incorporation of nodes with an alignment parallel to the neighboring nodes already residing in the crystal.

  4. Radial lean direct injection burner

    Science.gov (United States)

    Khan, Abdul Rafey; Kraemer, Gilbert Otto; Stevenson, Christian Xavier

    2012-09-04

    A burner for use in a gas turbine engine includes a burner tube having an inlet end and an outlet end; a plurality of air passages extending axially in the burner tube configured to convey air flows from the inlet end to the outlet end; a plurality of fuel passages extending axially along the burner tube and spaced around the plurality of air passage configured to convey fuel from the inlet end to the outlet end; and a radial air swirler provided at the outlet end configured to direct the air flows radially toward the outlet end and impart swirl to the air flows. The radial air swirler includes a plurality of vanes to direct and swirl the air flows and an end plate. The end plate includes a plurality of fuel injection holes to inject the fuel radially into the swirling air flows. A method of mixing air and fuel in a burner of a gas turbine is also provided. The burner includes a burner tube including an inlet end, an outlet end, a plurality of axial air passages, and a plurality of axial fuel passages. The method includes introducing an air flow into the air passages at the inlet end; introducing a fuel into fuel passages; swirling the air flow at the outlet end; and radially injecting the fuel into the swirling air flow.

  5. AP1000 Design for Security

    International Nuclear Information System (INIS)

    Long, L.B.; Cummins, W.E.; Winters, J.W.

    2006-01-01

    Nuclear power plants are protected from potential security threats through a combination of robust structures around the primary system and other vital equipment, security systems and equipment, and defensive strategy. The overall objective for nuclear power plant security is to protect public health and safety by ensuring that attacks or sabotage do not challenge the ability to safely shutdown the plant or protect from radiological releases. In addition, plants have systems, features and operational strategies to cope with external conditions, such as loss of offsite power, which could be created as part of an attack. Westinghouse considered potential security threats during design of the AP1000 PWR. The differences in plant configuration, safety system design, and safe shutdown equipment between existing plants and AP1000 affect potential vulnerabilities. This paper provides an evaluation of AP1000 with respect to vulnerabilities to security threats. The AP1000 design differs from the design of operating PWRs in the US in the configuration and the functional requirements for safety systems. These differences are intentional departures from conventional PWR designs which simplify plant design and enhance overall safety. The differences between the AP1000 PWR and conventional PWRs can impact vulnerabilities to security threats. The NRC addressed security concerns as part of their reviews for AP1000 Design Certification, and did not identify any security issues of concern. However, much of the detailed security design information for the AP1000 was deferred to the combined Construction and Operating License (COL) phase as many of the security issues are site-specific. Therefore, NRC review of security issues related to the AP1000 is not necessarily complete. Further, since the AP1000 plant design differs from existing PWRs, it is not obvious that the analyses and assessments prepared for existing plants also apply to the AP1000. We conclude that, overall, the AP1000

  6. Anomalies of radial and ulnar arteries

    Directory of Open Access Journals (Sweden)

    Rajani Singh

    Full Text Available Abstract During dissection conducted in an anatomy department of the right upper limb of the cadaver of a 70-year-old male, both origin and course of the radial and ulnar arteries were found to be anomalous. After descending 5.5 cm from the lower border of the teres major, the brachial artery anomalously bifurcated into a radial artery medially and an ulnar artery laterally. In the arm, the ulnar artery lay lateral to the median nerve. It followed a normal course in the forearm. The radial artery was medial to the median nerve in the arm and then, at the level of the medial epicondyle, it crossed from the medial to the lateral side of the forearm, superficial to the flexor muscles. The course of the radial artery was superficial and tortuous throughout the arm and forearm. The variations of radial and ulnar arteries described above were associated with anomalous formation and course of the median nerve in the arm. Knowledge of neurovascular anomalies are important for vascular surgeons and radiologists.

  7. Variations in the usage and composition of a radial cocktail during radial access coronary angiography procedures.

    LENUS (Irish Health Repository)

    Pate, G

    2011-10-01

    A survey was conducted of medication administered during radial artery cannulation for coronary angiography in 2009 in Ireland; responses were obtained for 15 of 20 centres, in 5 of which no radial access procedures were undertaken. All 10 (100%) centres which provided data used heparin and one or more anti-spasmodics; verapamil in 9 (90%), nitrate in 1 (10%), both in 2 (20%). There were significant variations in the doses used. Further work needs to be done to determine the optimum cocktail to prevent radial artery injury following coronary angiography.

  8. Design of radial reinforcement for prestressed concrete containments

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shen, E-mail: swang@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States); Munshi, Javeed A., E-mail: jamunshi@bechtel.com [Bechtel Power Corporation, 5275 Westview Drive, BP2-2C3, Frederick, MD 21703 (United States)

    2013-02-15

    Highlights: ► A rigorous formulae is proposed to calculate radial stress within prestressed concrete containments. ► The proposed method is validated by finite element analysis in an illustrative practical example. ► A partially prestressed condition is more critical than a fully prestressed condition for radial tension. ► Practical design consideration is provided for detailing of radial reinforcement. -- Abstract: Nuclear containments are critical components for safety of nuclear power plants. Failure can result in catastrophic safety consequences as a result of leakage of radiation. Prestressed concrete containments have been used in large nuclear power plants with significant design internal pressure. These containments are generally reinforced with prestressing tendons in the circumferential (hoop) and meridional (vertical) directions. The curvature effect of the tendons introduces radial tensile stresses in the concrete shell which are generally neglected in the design of such structures. It is assumed that such tensile radial stresses are small as such no radial reinforcement is provided for this purpose. But recent instances of significant delaminations in Crystal River Unit 3 in Florida have elevated the need for reevaluation of the radial tension issue in prestressed containment. Note that currently there are no well accepted industry standards for design and detailing of radial reinforcement. This paper discusses the issue of radial tension in prestressed cylindrical and dome shaped structures and proposes formulae to calculate radial stresses. A practical example is presented to illustrate the use of the proposed method which is then verified by using state of art finite element analysis. This paper also provides some practical design consideration for detailing of radial reinforcement in prestressed containments.

  9. Methods and apparatus for radially compliant component mounting

    Science.gov (United States)

    Bulman, David Edward [Cincinnati, OH; Darkins, Jr., Toby George; Stumpf, James Anthony [Columbus, IN; Schroder, Mark S [Greenville, SC; Lipinski, John Joseph [Simpsonville, SC

    2012-03-27

    Methods and apparatus for a mounting assembly for a liner of a gas turbine engine combustor are provided. The combustor includes a combustor liner and a radially outer annular flow sleeve. The mounting assembly includes an inner ring surrounding a radially outer surface of the liner and including a plurality of axially extending fingers. The mounting assembly also includes a radially outer ring coupled to the inner ring through a plurality of spacers that extend radially from a radially outer surface of the inner ring to the outer ring.

  10. Fuel-steel mixing and radial mesh effects in power excursion simulations

    International Nuclear Information System (INIS)

    Chen, X.-N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Vezzoni, B.; Li, R.; Maschek, W.; Kiefhaber, E.

    2016-01-01

    Highlights: • Fuel-steel mixing and radial mesh effects are significant on power excursion. • The earliest power peak is reduced and retarded by these two effects. • Unprotected loss of coolant transients in ESFR core are calculated. - Abstract: This paper deals with SIMMER-III once-through simulations of the earliest power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered originally in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is −2 $, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly (SA) coarse mesh treatment, where a simultaneous sodium boiling onset in all SAs belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped “homogenization” reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the earliest power excursion after the sodium boiling.

  11. Computer model analysis of the radial artery pressure waveform.

    Science.gov (United States)

    Schwid, H A; Taylor, L A; Smith, N T

    1987-10-01

    Simultaneous measurements of aortic and radial artery pressures are reviewed, and a model of the cardiovascular system is presented. The model is based on resonant networks for the aorta and axillo-brachial-radial arterial system. The model chosen is a simple one, in order to make interpretation of the observed relationships clear. Despite its simplicity, the model produces realistic aortic and radial artery pressure waveforms. It demonstrates that the resonant properties of the arterial wall significantly alter the pressure waveform as it is propagated from the aorta to the radial artery. Although the mean and end-diastolic radial pressures are usually accurate estimates of the corresponding aortic pressures, the systolic pressure at the radial artery is often much higher than that of the aorta due to overshoot caused by the resonant behavior of the radial artery. The radial artery dicrotic notch is predominantly dependent on the axillo-brachial-radial arterial wall properties, rather than on the aortic valve or peripheral resistance. Hence the use of the radial artery dicrotic notch as an estimate of end systole is unreliable. The rate of systolic upstroke, dP/dt, of the radial artery waveform is a function of many factors, making it difficult to interpret. The radial artery waveform usually provides accurate estimates for mean and diastolic aortic pressures; for all other measurements it is an inadequate substitute for the aortic pressure waveform. In the presence of low forearm peripheral resistance the mean radial artery pressure may significantly underestimate the mean aortic pressure, as explained by a voltage divider model.

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.)

  17. MR accuracy and arthroscopic incidence of meniscal radial tears

    Energy Technology Data Exchange (ETDEWEB)

    Magee, Thomas; Shapiro, Marc; Williams, David [Department of Radiology, Neuroimaging Institute, 27 East Hibiscus Blvd., Melbourne, FL 32901 (United States)

    2002-12-01

    A meniscal radial tear is a vertical tear that involves the inner meniscal margin. The tear is most frequent in the middle third of the lateral meniscus and may extend outward in any direction. We report (1) the arthroscopic incidence of radial tears, (2) MR signs that aid in the detection of radial tears and (3) our prospective accuracy in detection of radial tears. Design and patients. Three musculoskeletal radiologists prospectively read 200 consecutive MR examinations of the knee that went on to arthroscopy by one orthopedic surgeon. MR images were assessed for location and MR characteristics of radial tears. MR criteria used for diagnosis of a radial tear were those outlined by Tuckman et al.: truncation, abnormal morphology and/or lack of continuity or absence of the meniscus on one or more MR images. An additional criterion used was abnormal increased signal in that area on fat-saturated proton density or T2-weighted coronal and sagittal images. Prospective MR readings were correlated with the arthroscopic findings.Results. Of the 200 consecutive knee arthroscopies, 28 patients had radial tears reported arthroscopically (14% incidence). MR readings prospectively demonstrated 19 of the 28 radial tears (68% sensitivity) when the criteria for diagnosis of a radial tear were truncation or abnormal morphology of the meniscus. With the use of the additional criterion of increased signal in the area of abnormal morphology on fat-saturated T2-weighted or proton density weighted sequences, the prospective sensitivity was 25 of 28 radial tears (89% sensitivity). There were no radial tears described in MR reports that were not demonstrated on arthroscopy (i.e., there were no false positive MR readings of radial tears in these 200 patients). Radial tears are commonly seen at arthroscopy. There was a 14% incidence in this series of 200 patients who underwent arthroscopy. Prospective detection of radial tears was 68% as compared with arthroscopy when the criteria as

  18. MR accuracy and arthroscopic incidence of meniscal radial tears

    International Nuclear Information System (INIS)

    Magee, Thomas; Shapiro, Marc; Williams, David

    2002-01-01

    A meniscal radial tear is a vertical tear that involves the inner meniscal margin. The tear is most frequent in the middle third of the lateral meniscus and may extend outward in any direction. We report (1) the arthroscopic incidence of radial tears, (2) MR signs that aid in the detection of radial tears and (3) our prospective accuracy in detection of radial tears. Design and patients. Three musculoskeletal radiologists prospectively read 200 consecutive MR examinations of the knee that went on to arthroscopy by one orthopedic surgeon. MR images were assessed for location and MR characteristics of radial tears. MR criteria used for diagnosis of a radial tear were those outlined by Tuckman et al.: truncation, abnormal morphology and/or lack of continuity or absence of the meniscus on one or more MR images. An additional criterion used was abnormal increased signal in that area on fat-saturated proton density or T2-weighted coronal and sagittal images. Prospective MR readings were correlated with the arthroscopic findings.Results. Of the 200 consecutive knee arthroscopies, 28 patients had radial tears reported arthroscopically (14% incidence). MR readings prospectively demonstrated 19 of the 28 radial tears (68% sensitivity) when the criteria for diagnosis of a radial tear were truncation or abnormal morphology of the meniscus. With the use of the additional criterion of increased signal in the area of abnormal morphology on fat-saturated T2-weighted or proton density weighted sequences, the prospective sensitivity was 25 of 28 radial tears (89% sensitivity). There were no radial tears described in MR reports that were not demonstrated on arthroscopy (i.e., there were no false positive MR readings of radial tears in these 200 patients). Radial tears are commonly seen at arthroscopy. There was a 14% incidence in this series of 200 patients who underwent arthroscopy. Prospective detection of radial tears was 68% as compared with arthroscopy when the criteria as

  19. 47 CFR 76.1000 - Definitions.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Definitions. 76.1000 Section 76.1000 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) BROADCAST RADIO SERVICES MULTICHANNEL VIDEO AND CABLE TELEVISION SERVICE Competitive Access to Cable Programming § 76.1000 Definitions. As used in this...

  20. 7 CFR 3017.1000 - Respondent.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 15 2010-01-01 2010-01-01 false Respondent. 3017.1000 Section 3017.1000 Agriculture Regulations of the Department of Agriculture (Continued) OFFICE OF THE CHIEF FINANCIAL OFFICER, DEPARTMENT OF AGRICULTURE GOVERNMENTWIDE DEBARMENT AND SUSPENSION (NONPROCUREMENT) Definitions § 3017.1000 Respondent...

  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. Radial velocities of RR Lyrae stars

    International Nuclear Information System (INIS)

    Hawley, S.L.; Barnes, T.G. III

    1985-01-01

    283 spectra of 57 RR Lyrae stars have been obtained using the 2.1-m telescope at McDonald Observatory. Radial velocities were determined using a software cross-correlation technique. New mean radial velocities were determined for 46 of the stars. 11 references

  3. Seamless Heterogeneous 3D Tessellation via DWT Domain Smoothing and Mosaicking

    Directory of Open Access Journals (Sweden)

    Gilles Gesquière

    2010-01-01

    Full Text Available With todays geobrowsers, the tessellations are far from being smooth due to a variety of reasons: the principal being the light difference and resolution heterogeneity. Whilst the former has been extensively dealt with in the literature through classic mosaicking techniques, the latter has got little attention. We focus on this latter aspect and present two DWT domain methods to seamlessly stitch tiles of heterogeneous resolutions. The first method is local in that each of the tiles that constitute the view, is subjected to one of the three context-based smoothing functions proposed for horizontal, vertical, and radial smoothing, depending on its localization in the tessellation. These functions are applied at the DWT subband level and followed by an inverse DWT to give a smoothened tile. In the second method, though we assume the same tessellation scenario, the view field is thought to be of a sliding window which may contain parts of the tiles from the heterogeneous tessellation. The window is refined in the DWT domain through mosaicking and smoothing followed by a global inverse DWT. Rather than the traditional sense, the mosaicking employed over here targets the heterogeneous resolution. Perceptually, this second method has shown better results than the first one. The methods have been successfully applied to practical examples of both the texture and its corresponding DEM for seamless 3D terrain visualization.

  4. 40 CFR 725.1000 - Scope.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Scope. 725.1000 Section 725.1000 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC SUBSTANCES CONTROL ACT REPORTING....1000 Scope. This subpart identifies uses of microorganisms which EPA has determined to be significant...

  5. 40 CFR 51.1000 - Definitions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 2 2010-07-01 2010-07-01 false Definitions. 51.1000 Section 51.1000 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS REQUIREMENTS FOR... Ambient Air Quality Standards § 51.1000 Definitions. The following definitions apply for purposes of this...

  6. 7 CFR 760.1000 - Applicability.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 7 2010-01-01 2010-01-01 false Applicability. 760.1000 Section 760.1000 Agriculture Regulations of the Department of Agriculture (Continued) FARM SERVICE AGENCY, DEPARTMENT OF AGRICULTURE... Catfish Grant Programs § 760.1000 Applicability. (a) This subpart establishes the terms and conditions...

  7. 46 CFR 154.1000 - Applicability.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Applicability. 154.1000 Section 154.1000 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY STANDARDS FOR... § 154.1000 Applicability. Sections 154.1005 through 154.1020 apply to flammable cargo and ammonia...

  8. Radial MR images of the knee

    International Nuclear Information System (INIS)

    Hewes, R.C.; Miller, T.R.

    1988-01-01

    To profile optimally each portion of the meniscus, the authors use the multiangle, multisection feature of a General Electric SIGNA 1.5-T imager to produce radial images centered on each meniscus. A total of 12-15 sections are imaged at 10 0 -15 0 intervals of each meniscus, yielding perpendicular images of the entire meniscus, comparable with the arthrographic tangential views. The authors review their technique and demonstrate correlation cases between the radial gradient recalled acquisition in a steady state sequences, sagittal and coronal MR images, and arthrograms. Radial images should be a routine part of knee MR imaging

  9. 21 CFR 866.4800 - Radial immunodiffusion plate.

    Science.gov (United States)

    2010-04-01

    ...) MEDICAL DEVICES IMMUNOLOGY AND MICROBIOLOGY DEVICES Immunology Laboratory Equipment and Reagents § 866.4800 Radial immunodiffusion plate. (a) Identification. A radial immunodiffusion plate for clinical use...

  10. Astrocyte morphology, heterogeneity and density in the developing African Giant Rat (Cricetomys gambianus

    Directory of Open Access Journals (Sweden)

    James Olukayode Olopade

    2015-05-01

    Full Text Available Astrocyte morphologies and heterogeneity were described in male African giant rats (AGR (Cricetomys gambianus, Waterhouse across three age groups (5 neonates, 5 juveniles and 5 adults using Silver impregnation method and immunohistochemistry against glia fibrillary acidic protein (GFAP. Immunopositive cell signaling, cell size and population were least in neonates, followed by adults and juveniles respectively. In neonates, astrocyte processes were mostly detected within the glia limitans of the mid and hind brain; their cell bodies measuring 32±4.8 µm in diameter against 91±5.4µm and 75± 1.9µm in juveniles and adults respectively. Astrocyte heterogeneity in juvenile and adult groups revealed eight subtypes to include fibrous astrocytes chiefly in the corpus callosum and brain stem, protoplasmic astrocytes in the cortex and dentate gyrus (DG; radial glia were found along the olfactory bulb (OB and subventricular zone (SVZ; velate astrocytes were mainly found in the cerebellum and hippocampus; marginal astrocytes close to the pia mater; Bergmann glia in the molecular layer of the cerebellum; perivascular and periventricular astrocytes in the cortex and third ventricle respectively. Cell counts from twelve anatomical regions of the brain were significantly higher in juveniles than in adults (p≤0.01 using unpaired student t-test in the cerebral cortex, pia, corpus callosum, rostral migratory stream (RMS, DG and cerebellum. Highest astrocyte count was found in the DG, while the least count was in the brain stem and sub cortex. Astrocytes along the periventricular layer of the OB are believed to be part of the radial glia system that transport newly formed cells towards the hippocampus and play roles in neurogenesis migration and homeostasis in the AGR. Therefore, astrocyte heterogeneity was examined across age groups in the AGR to determine whether age influences astrocytes population in different regions of the AGR brain and discuss

  11. 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

  12. 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

  13. 29 CFR 98.1000 - Respondent.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 1 2010-07-01 2010-07-01 true Respondent. 98.1000 Section 98.1000 Labor Office of the Secretary of Labor GOVERNMENTWIDE DEBARMENT AND SUSPENSION (NONPROCUREMENT) Definitions § 98.1000 Respondent. Respondent means a person against whom an agency has initiated a debarment or suspension action. ...

  14. 33 CFR 151.1000 - Purpose.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Purpose. 151.1000 Section 151.1000 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) POLLUTION... Transportation of Municipal and Commercial Waste § 151.1000 Purpose. The purpose of this subpart is to implement...

  15. 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)

  16. 21 CFR 1000.1 - General.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false General. 1000.1 Section 1000.1 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) RADIOLOGICAL HEALTH GENERAL General Provisions § 1000.1 General. References in this subchapter J to regulatory sections of the...

  17. Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

    International Nuclear Information System (INIS)

    Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

    2017-01-01

    Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

  18. 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)

  19. 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

  20. Stirling Engine With Radial Flow Heat Exchangers

    Science.gov (United States)

    Vitale, N.; Yarr, George

    1993-01-01

    Conflict between thermodynamical and structural requirements resolved. In Stirling engine of new cylindrical configuration, regenerator and acceptor and rejector heat exchangers channel flow of working gas in radial direction. Isotherms in regenerator ideally concentric cylinders, and gradient of temperature across regenerator radial rather than axial. Acceptor and rejector heat exchangers located radially inward and outward of regenerator, respectively. Enables substantial increase in power of engine without corresponding increase in diameter of pressure vessel.

  1. 25 CFR 1000.1 - Authority.

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 2 2010-04-01 2010-04-01 false Authority. 1000.1 Section 1000.1 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL... § 1000.1 Authority. This part is prepared and issued by the Secretary of the Interior under the...

  2. Comparison between ultrasound guided technique and digital palpation technique for radial artery cannulation in adult patients: An updated meta-analysis of randomized controlled trials.

    Science.gov (United States)

    Bhattacharjee, Sulagna; Maitra, Souvik; Baidya, Dalim K

    2018-03-22

    Possible advantages and risks associated with ultrasound guided radial artery cannulation in-comparison to digital palpation guided method in adult patients are not fully known. We have compared ultrasound guided radial artery cannulation with digital palpation technique in this meta-analysis. Meta-analysis of randomized controlled trials. Trials conducted in operating room, emergency department, cardiac catheterization laboratory. PubMed and Cochrane Central Register of Controlled Trials (CENTRAL) were searched (from 1946 to 20th November 2017) to identify prospective randomized controlled trials in adult patients. Two-dimensional ultrasound guided radial artery catheterization versus digital palpation guided radial artery cannulation. Overall cannulation success rate, first attempt success rate, time to cannulation and mean number of attempts to successful cannulation. Odds ratio (OR) and standardized mean difference (SMD) or mean difference (MD) with 95% confidence interval (CI) were calculated for categorical and continuous variables respectively. Data of 1895 patients from 10 studies have been included in this meta- analysis. Overall cannulation success rate was similar between ultrasound guided technique and digital palpation [OR (95% CI) 2.01 (1.00, 4.06); p = 0.05]. Ultrasound guided radial artery cannulation is associated with higher first attempt success rate of radial artery cannulation in comparison to digital palpation [OR (95% CI) 2.76 (186, 4.10); p guided technique with palpation technique. Radial artery cannulation by ultrasound guidance may increase the first attempt success rate but not the overall cannulation success when compared to digital palpation technique. However, results of this meta-analysis should be interpreted with caution due presence of heterogeneity. Copyright © 2018. Published by Elsevier Inc.

  3. Radial wedge flange clamp

    Science.gov (United States)

    Smith, Karl H.

    2002-01-01

    A radial wedge flange clamp comprising a pair of flanges each comprising a plurality of peripheral flat wedge facets having flat wedge surfaces and opposed and mating flat surfaces attached to or otherwise engaged with two elements to be joined and including a series of generally U-shaped wedge clamps each having flat wedge interior surfaces and engaging one pair of said peripheral flat wedge facets. Each of said generally U-shaped wedge clamps has in its opposing extremities apertures for the tangential insertion of bolts to apply uniform radial force to said wedge clamps when assembled about said wedge segments.

  4. 21 CFR 573.1000 - Verxite.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Verxite. 573.1000 Section 573.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS, FEEDS... Listing § 573.1000 Verxite. The food additive verxite may be safely used in animal feed in accordance with...

  5. 25 CFR 11.1000 - Complaint.

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false Complaint. 11.1000 Section 11.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR LAW AND ORDER COURTS OF INDIAN OFFENSES AND LAW AND ORDER CODE Juvenile Offender Procedure § 11.1000 Complaint. A complaint must be filed by a law enforcement officer or...

  6. Radial pattern of nuclear decay processes

    International Nuclear Information System (INIS)

    Iskra, W.; Mueller, M.; Rotter, I.; Technische Univ. Dresden

    1994-05-01

    At high level density of nuclear states, a separation of different time scales is observed (trapping effect). We calculate the radial profile of partial widths in the framework of the continuum shell model for some 1 - resonances with 2p-2h nuclear structure in 16 O as a function of the coupling strength to the continuum. A correlation between the lifetime of a nuclear state and the radial profile of the corresponding decay process is observed. We conclude from our numerical results that the trapping effect creates structures in space and time characterized by a small radial extension and a short lifetime. (orig.)

  7. AC magnetic transport on heterogeneous ferromagnetic wires and tubes

    International Nuclear Information System (INIS)

    Sinnecker, J.P.; Pirota, K.R.; Knobel, M.; Kraus, L.

    2002-01-01

    The AC current density radial distribution is calculated on heterogeneous composite materials with cylindrical geometry. The composites have an inner core and thin outer shell that can be either from the same material (homogenous material like simple wires) or from different materials with different physical properties. The case in which a non-magnetic inner core is surrounded by a magnetic layer, like electrodeposited wires, is mainly studied. The effect of frequency and applied magnetic field is simulated. The current density distribution as a function of frequency and applied field, as well as the total current over the inner core and outer shells are calculated. The results agree substantially well with the experimentally observed data for simple electrodeposited wires

  8. Intraluminal milrinone for dilation of the radial artery graft.

    Science.gov (United States)

    García-Rinaldi, R; Soltero, E R; Carballido, J; Mojica, J

    1999-01-01

    There is renewed interest in the use of the radial artery as a conduit for coronary artery bypass surgery. The radial artery is, however, a very muscular artery, prone to vasospasm. Milrinone, a potent vasodilator, has demonstrated vasodilatory properties superior to those of papaverine. In this report, we describe our technique of radial artery harvesting and the adjunctive use of intraluminal milrinone as a vasodilator in the preparation of this conduit for coronary artery bypass grafting. We have used these techniques in 25 patients who have undergone coronary artery bypass grafting using the radial artery. No hand ischemic complications have been observed in this group. Intraluminal milrinone appears to dilate and relax the radial artery, rendering this large conduit spasm free and very easy to use. We recommend the skeletonization technique for radial artery harvesting and the use of intraluminal milrinone as a radial artery vasodilator in routine myocardial revascularization. PMID:10524740

  9. 30 CFR 250.1000 - General requirements.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 2 2010-07-01 2010-07-01 false General requirements. 250.1000 Section 250.1000... OPERATIONS IN THE OUTER CONTINENTAL SHELF Pipelines and Pipeline Rights-of-Way § 250.1000 General....1001, must meet the requirements in §§ 250.1000 through 250.1008. (2) A pipeline right-of-way grant...

  10. 7 CFR 1000.54 - Equivalent price.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true Equivalent price. 1000.54 Section 1000.54 Agriculture... Prices § 1000.54 Equivalent price. If for any reason a price or pricing constituent required for computing the prices described in § 1000.50 is not available, the market administrator shall use a price or...

  11. Nuclear design manual for generation of cross section and heterogeneous formfunction for CASMO-3/MASTER

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Ho; Cho, Byung Oh; Song, Jae Seong; Lee, Chung Chan

    1996-12-01

    A three-dimensional reactor core simulation code, MASTER, has been developed as a part of the ADONIS project in KAERI. CASMO-3 prepares various two-group cross sections for the constituents of a reactor core such as fuel assembly, radial and axial reflectors, control rod and detector for MASTER. This report includes the standard design procedure for generation of two-group cross sections and heterogeneous formfunction by CASMO-3/FORM for MASTER. (author). 16 refs., 16 tabs., 12 figs.

  12. Channeling of protons through radial deformed carbon nanotubes

    Energy Technology Data Exchange (ETDEWEB)

    Borka Jovanović, V., E-mail: vborka@vinca.rs [Atomic Physics Laboratory (040), Vinča Institute of Nuclear Sciences, University of Belgrade, P.O. Box 522, 11001 Belgrade (Serbia); Borka, D. [Atomic Physics Laboratory (040), Vinča Institute of Nuclear Sciences, University of Belgrade, P.O. Box 522, 11001 Belgrade (Serbia); Galijaš, S.M.D. [Faculty of Physics, University of Belgrade, P.O. Box 368, 11001 Belgrade (Serbia)

    2017-05-18

    Highlights: • For the first time we presented theoretically obtained distributions of channeled protons with radially deformed SWNT. • Our findings indicate that influence of the radial deformation is very strong and it should not be omitted in simulations. • We show that the spatial and angular distributions depend strongly of level of radial deformation of nanotube. • Our obtained results can be compared with measured distributions to reveal the presence of various types of defects in SWNT. - Abstract: In this paper we have presented a theoretical investigation of the channeling of 1 GeV protons with the radial deformed (10, 0) single-wall carbon nanotubes (SWNTs). We have calculated channeling potential within the deformed nanotubes. For the first time we presented theoretically obtained spatial and angular distributions of channeled protons with radially deformed SWNT. We used a Monte Carlo (MC) simulation technique. We show that the spatial and angular distributions depend strongly of level of radial deformation of nanotube. These results may be useful for nanotube characterization and production and guiding of nanosized ion beams.

  13. 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)

  14. 24 CFR 1000.28 - May a self-governance Indian tribe be exempted from the applicability of § 1000.26?

    Science.gov (United States)

    2010-04-01

    ... be exempted from the applicability of § 1000.26? 1000.28 Section 1000.28 Housing and Urban... ACTIVITIES General § 1000.28 May a self-governance Indian tribe be exempted from the applicability of § 1000... and systems meet or exceed the comparable requirements of § 1000.26. For purposes of this section, a...

  15. The effect of material heterogeneity in curved composite beams for use in aircraft structures

    Science.gov (United States)

    Otoole, Brendan J.; Santare, Michael H.

    1992-01-01

    A design tool is presented for predicting the effect of material heterogeneity on the performance of curved composite beams for use in aircraft fuselage structures. Material heterogeneity can be induced during processes such as sheet forming and stretch forming of thermoplastic composites. This heterogeneity can be introduced in the form of fiber realignment and spreading during the manufacturing process causing a gradient in material properties in both the radial and tangential directions. The analysis procedure uses a separate two-dimensional elasticity solution for the stresses in the flanges and web sections of the beam. The separate solutions are coupled by requiring the forces and displacements match at the section boundaries. Analysis is performed for curved beams loaded in pure bending and uniform pressure. The beams can be of any general cross-section such as a hat, T-, I-, or J-beam. Preliminary results show that geometry of the beam dictates the effect of heterogeneity on performance. Heterogeneity plays a much larger role in beams with a small average radius to depth ratio, R/t, where R is the average radius of the beam and t is the difference between the inside and outside radius. Results of the analysis are in the form of stresses and displacements, and they are compared to both mechanics of materials and numerical solutions obtained using finite element analysis.

  16. AP1000. The PWR revisited

    International Nuclear Information System (INIS)

    Gaio, P.

    2006-01-01

    The distinguishing features of Westinghouse's AP1000 advanced passive pressurized water reactor are highlighted. In particular, the AP1000's passive safety features are described as well as their implications for simplifying the design, construction, and operation of this design compared to currently operating plants, and significantly increasing safety margins over current plants as well. The AP1000 design specifically incorporates the knowledge acquired from the substantial accumulation of power reactor operating experience and benefits from the application of the Probabilistic Risk Assessment in the design process itself. The AP1000 design has been certified by the US Nuclear Regulatory Commission under its new rules for licensing new nuclear plants, 10 CFR Part 52, and is the subject of six combined Construction and Operating License applications now being developed. Currently the AP1000 design is being assessed against the EUR Rev C requirements for new nuclear power plants in Europe. (author)

  17. 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.

  18. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  19. 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

  20. 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.

  1. Interactions between Radial Electric Field, Transport and Structure in Helical Plasmas

    International Nuclear Information System (INIS)

    Ida, Katsumi and others

    2006-01-01

    Control of the radial electric field is considered to be important in helical plasmas, because the radial electric field and its shear are expected to reduce neoclassical and anomalous transport, respectively. Particle and heat transport, that determines the radial structure of density and electron profiles, sensitive to the structure of radial electric field. On the other hand, the radial electric field itself is determined by the plasma parameters. In general, the sign of the radial electric field is determined by the plasma collisionality, while the magnitude of the radial electric field is determined by the temperature and/or density gradients. Therefore the structure of radial electric field and temperature and density are strongly coupled through the particle and heat transport and formation mechanism of radial electric field. Interactions between radial electric field, transport and structure in helical plasmas is discussed based on the experiments on Large Helical Device

  2. Anomalous Medial Branch of Radial Artery: A Rare Variant

    Directory of Open Access Journals (Sweden)

    Surbhi Wadhwa

    2016-10-01

    Full Text Available Radial artery is an important consistent vessel of the upper limb. It is a useful vascular access site for coronary procedures and its reliable anatomy has resulted in an elevation of radial forearm flaps for reconstructive surgeries of head and neck. Technical failures, in both the procedures, are mainly due to anatomical variations, such as radial loops, ectopic radial arteries or tortuosity in the vessel. We present a rare and a unique anomalous medial branch of the radial artery spiraling around the flexor carpi radialis muscle in the forearm with a high rising superficial palmar branch of radial artery. Developmentally it probably is a remanent of the normal pattern of capillary vessel maintenance and regression. Such a case is of importance for reconstructive surgeons and coronary interventionists, especially in view of its unique medial and deep course.

  3. Fuel radial design using Path Relinking; Diseno radial de combustible usando Path Relinking

    Energy Technology Data Exchange (ETDEWEB)

    Campos S, Y. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2007-07-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  4. Clinical and Radiographic Outcomes of Unipolar and Bipolar Radial Head Prosthesis in Patients with Radial Head Fracture: A Systemic Review and Meta-Analysis.

    Science.gov (United States)

    Chen, Hongwei; Wang, Ziyang; Shang, Yongjun

    2018-06-01

    To compare clinical outcomes of unipolar and bipolar radial head prosthesis in the treatment of patients with radial head fracture. Medline, Cochrane, EMBASE, Google Scholar databases were searched until April 18, 2016 using the following search terms: radial head fracture, elbow fracture, radial head arthroplasty, implants, prosthesis, unipolar, bipolar, cemented, and press-fit. Randomized controlled trials, retrospective, and cohort studies were included. The Mayo elbow performance score (MEPS), disabilities of the arm, shoulder, and hand (DASH) score, radiologic assessment, ROM, and grip strength following elbow replacement were similar between prosthetic devices. The pooled mean excellent/good ranking of MEPS was 0.78 for unipolar and 0.73 for bipolar radial head arthroplasty, and the pooled mean MEPS was 86.9 and 79.9, respectively. DASH scores for unipolar and bipolar prosthesis were 19.0 and 16.3, respectively. Range of motion outcomes were similar between groups, with both groups have comparable risk of flexion arc, flexion, extension deficit, rotation arc, pronation, and supination (p values bipolar prosthesis). However, bipolar radial head prosthesis was associated with an increased chance of heterotopic ossification and lucency (p values ≤0.049) while unipolar prosthesis was not (p values ≥0.088). Both groups had risk for development of capitellar osteopenia or erosion/wear (p values ≤0.039). Unipolar and bipolar radial head prostheses were similar with respect to clinical outcomes. Additional comparative studies are necessary to further compare different radial head prostheses used to treat radial head fracture.

  5. 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

  6. 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)

  7. 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)

  8. Vitreous veils and radial lattice in Marshall syndrome.

    Science.gov (United States)

    Brubaker, Jacob W; Mohney, Brian G; Pulido, Jose S; Babovic-Vuksanovic, Dusica

    2008-12-01

    To report the findings of membranous vitreous veils and radial lattice in a child with Marshall syndrome. Observational case report. Retrospective review of medical records and fundus photograph of a 6-year-old boy with Marshall syndrome. Vitreoretinal findings were significant for bilateral membranous vitreous veils and radial lattice degeneration. This case demonstrates the occurrence of vitreous veils and radial lattice degeneration in patients with Marshall syndrome.

  9. Long-Term Follow-up of Modular Metallic Radial Head Replacement: Commentary on an article by Jonathan P. Marsh, MD, FRCSC, et al.: "Radial Head Fractures Treated with Modular Metallic Radial Head Replacement: Outcomes at a Mean Follow-up of Eight Years".

    OpenAIRE

    Mansat, Pierre

    2016-01-01

    Radial head arthroplasty is used to stabilize the joint after a complex acute radial head fracture that is not amenable for fixation or to treat sequelae of radial head fractures. Most of the currently used radial head prostheses are metallic monoblock implants that are not consistently adaptable and raise technical challenges since their implantation requires lateral elbow subluxation. Metallic modular radial head arthroplasty implants available in various head and stem sizes have been devel...

  10. 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)

  11. 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

  12. Radial head fracture associated with posterior interosseous nerve injury

    Directory of Open Access Journals (Sweden)

    Bernardo Barcellos Terra

    Full Text Available ABSTRACT Fractures of the radial head and radial neck correspond to 1.7-5.4% of all fractures and approximately 30% may present associated injuries. In the literature, there are few reports of radial head fracture with posterior interosseous nerve injury. This study aimed to report a case of radial head fracture associated with posterior interosseous nerve injury. CASE REPORT: A male patient, aged 42 years, sought medical care after falling from a skateboard. The patient related pain and limitation of movement in the right elbow and difficulty to extend the fingers of the right hand. During physical examination, thumb and fingers extension deficit was observed. The wrist extension showed a slight radial deviation. After imaging, it became evident that the patient had a fracture of the radial head that was classified as grade III in the Mason classification. The patient underwent fracture fixation; at the first postoperative day, thumb and fingers extension was observed. Although rare, posterior interosseous nerve branch injury may be associated with radial head fractures. In the present case, the authors believe that neuropraxia occurred as a result of the fracture hematoma and edema.

  13. Radial Field Piezoelectric Diaphragms

    Science.gov (United States)

    Bryant, R. G.; Effinger, R. T., IV; Copeland, B. M., Jr.

    2002-01-01

    A series of active piezoelectric diaphragms were fabricated and patterned with several geometrically defined Inter-Circulating Electrodes "ICE" and Interdigitated Ring Electrodes "ICE". When a voltage potential is applied to the electrodes, the result is a radially distributed electric field that mechanically strains the piezoceramic along the Z-axis (perpendicular to the applied electric field). Unlike other piezoelectric bender actuators, these Radial Field Diaphragms (RFDs) strain concentrically yet afford high displacements (several times that of the equivalent Unimorph) while maintaining a constant circumference. One of the more intriguing aspects is that the radial strain field reverses itself along the radius of the RFD while the tangential strain remains relatively constant. The result is a Z-deflection that has a conical profile. This paper covers the fabrication and characterization of the 5 cm. (2 in.) diaphragms as a function of poling field strength, ceramic thickness, electrode type and line spacing, as well as the surface topography, the resulting strain field and displacement as a function of applied voltage at low frequencies. The unique features of these RFDs include the ability to be clamped about their perimeter with little or no change in displacement, the environmentally insulated packaging, and a highly repeatable fabrication process that uses commodity materials.

  14. K-KIDS: K Dwarfs and Their Companions. First Results from Radial Velocity Survey with CHIRON Spectrograph

    Science.gov (United States)

    Paredes, Leonardo; Henry, Todd; Nusdeo, Daniel; Winters, J.; Dincer, Tolga

    2018-01-01

    We present the K-KIDS project, an effort to survey a large sample of K dwarfs and their companions, the KIDS. We are observing a carefully vetted equatorial sample (DEC = -30 to +30) of more than 1000 K dwarfs within 50 pc to make a comprehensive assessment of stellar, substellar and planetary companions with separations of 0.1 to 10,000 AU.The initial sample of 1048 stars has been compiled using astrometric data from Hipparcos and photometric data from Tycho-2 and 2MASS. Four different imaging and spectroscopic surveys are underway. Here we present the strategy and initial results for our high-precision radial velocity survey for the closest companions using the CHIRON spectrograph on the CTIO/SMARTS 1.5m telescope. Individual measurements with CHIRON at R = 80,000 using ThAr wavelength calibration, indicate that for K dwarf radial velocity standards with V = 5.8, 7.0 and 8.0 yield precisions over 6 weeks of observing of 7.4 m/s, 9.8 m/s and 5.7 m/s. In the first two months, a core sample of 42 K dwarfs, including carefully selected calibration systems as well as previously unobserved stars, was observed every few nights to detect the radial velocity signals of close companions. In our calibration stellar systems, we have confirmed the suitability of CHIRON for our studies, by having found periodic radial velocity perturbations consistent with hot Jupiter and stellar companions previously detected. This set forms the foundation of our one-year survey of 100 K dwarfs with magnitudes as faint as V = 11.5, for which we should detect companions with masses as low as Jupiter.In light of the promising performance and efficiency of the CHIRON spectrograph for a long-term radial velocity survey, we have expanded our initial sample using Gaia Data Release 1 to 1824 K dwarfs within 50 pc. Ultimately, the combination of all four surveys will provide an unprecedented portrait of K dwarfs and their kids.This effort has been supported by the NSF through grant AST-1517413, and

  15. Super cool X-1000 and Super cool Z-1000, two ice blockers, and their effect on vitrification/warming of mouse embryos.

    Science.gov (United States)

    Badrzadeh, H; Najmabadi, S; Paymani, R; Macaso, T; Azadbadi, Z; Ahmady, A

    2010-07-01

    To evaluate the survival and blastocyst formation rates of mouse embryos after vitrification/thaw process with different ice blocker media. We used X-1000 and Z-1000 separately and mixed using V-Kim, a closed vitrification system. Mouse embryos were vitrified using ethylene glycol based medium supplemented with Super cool X-1000 and/or Super cool Z-1000. Survival rates for the control, Super cool X-1000, Super cool Z-1000, and Super cool X-1000/Z-1000 groups were 74%, 72%, 68%, and 85% respectively, with no significant difference among experimental and control groups; however, a significantly higher survival rate was noticed in the Super cool X-1000/Z-1000 group when compared with the Super cool Z-1000 group. Blastocyst formation rates for the control, Super cool X-1000, Super cool Z-1000, and Super cool X-1000/Z-1000 groups were 71%, 66%, 65%, and 72% respectively. There was no significant difference in this rate among control and experimental groups. In a closed vitrification system, addition of ice blocker Super cool X-1000 to the vitrification solution containing Super cool Z-1000 may improve the embryo survival rate. We recommend combined ice blocker usage to optimize the vitrification outcome. Copyright (c) 2010 Elsevier Ireland Ltd. All rights reserved.

  16. 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

  17. Radial transport with perturbed magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Hazeltine, R. D. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States)

    2015-05-15

    It is pointed out that the viscosity coefficient describing radial transport of toroidal angular momentum is proportional to the second power of the gyro-radius—like the corresponding coefficients for particle and heat transport—regardless of any geometrical symmetry. The observation is widely appreciated, but worth emphasizing because some literature gives the misleading impression that asymmetry can allow radial moment transport in first-order.

  18. Radial transport with perturbed magnetic field

    International Nuclear Information System (INIS)

    Hazeltine, R. D.

    2015-01-01

    It is pointed out that the viscosity coefficient describing radial transport of toroidal angular momentum is proportional to the second power of the gyro-radius—like the corresponding coefficients for particle and heat transport—regardless of any geometrical symmetry. The observation is widely appreciated, but worth emphasizing because some literature gives the misleading impression that asymmetry can allow radial moment transport in first-order

  19. Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

    International Nuclear Information System (INIS)

    Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

    2015-01-01

    AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

  20. Rapid assessment of pulmonary gas transport with hyperpolarized 129Xe MRI using a 3D radial double golden-means acquisition with variable flip angles.

    Science.gov (United States)

    Ruppert, Kai; Amzajerdian, Faraz; Hamedani, Hooman; Xin, Yi; Loza, Luis; Achekzai, Tahmina; Duncan, Ian F; Profka, Harrilla; Siddiqui, Sarmad; Pourfathi, Mehrdad; Cereda, Maurizio F; Kadlecek, Stephen; Rizi, Rahim R

    2018-04-22

    To demonstrate the feasibility of using a 3D radial double golden-means acquisition with variable flip angles to monitor pulmonary gas transport in a single breath hold with hyperpolarized xenon-129 MRI. Hyperpolarized xenon-129 MRI scans with interleaved gas-phase and dissolved-phase excitations were performed using a 3D radial double golden-means acquisition in mechanically ventilated rabbits. The flip angle was either held fixed at 15 ° or 5 °, or it was varied linearly in ascending or descending order between 5 ° and 15 ° over a sampling interval of 1000 spokes. Dissolved-phase and gas-phase images were reconstructed at high resolution (32 × 32 × 32 matrix size) using all 1000 spokes, or at low resolution (22 × 22 × 22 matrix size) using 400 spokes at a time in a sliding-window fashion. Based on these sliding-window images, relative change maps were obtained using the highest mean flip angle as the reference, and aggregated pixel-based changes were tracked. Although the signal intensities in the dissolve-phase maps were mostly constant in the fixed flip-angle acquisitions, they varied significantly as a function of average flip angle in the variable flip-angle acquisitions. The latter trend reflects the underlying changes in observed dissolve-phase magnetization distribution due to pulmonary gas uptake and transport. 3D radial double golden-means acquisitions with variable flip angles provide a robust means for rapidly assessing lung function during a single breath hold, thereby constituting a particularly valuable tool for imaging uncooperative or pediatric patient populations. © 2018 International Society for Magnetic Resonance in Medicine.

  1. Impacts on human health from the coal and nuclear fuel cycles and other technologies associated with electric power generation and transmission

    International Nuclear Information System (INIS)

    Radford, E.P.

    1980-07-01

    The report evaluates major public health impacts of electric power generation and transmission associated with the nuclear fuel cycle and with coal use. Only existing technology is evaluated. For the nuclear cycle, effects of future use of fuel reprocessing and long-term radioactive waste disposal are briefly considered. The health effects of concern are those leading to definable human disease and injury. Health effects are scaled to numbers of persons and activities associated with a nominal 1000-megawatt electric plant fueled by either option. Comparison of the total health effects to the general public shows that the health risks from the coal cycle are about 50 times greater than for the nuclear cycle (coal, 0.7-3.7 major health effects per 1000 MWe per year; nuclear, 0.03-0.05 per 1000 MWe per year). For workers, these rates are higher. No evidence is found that electrical transmission contributes any health effects to the general public, except when broken power lines come in contact with people

  2. Present status and recent improvements of water chemistry at Russian VVER plants

    International Nuclear Information System (INIS)

    Mamet, V.; Yurmanov, V.

    2001-01-01

    Water chemistry is an important contributor to reliable plant operation, safety barrier integrity, plant component lifetime, radiation safety, environmental impact. Primary and secondary water chemistry guidelines of Russian VVER plants have been modified to meet the new safety standards. At present 14 VVER units of different generation are in operation at 5 Russian NPPs. There are eight 4-loop pressurised water reactors VVER-1000 (1000 MWe) and six 6-loop pressurised water reactors VVER-440 (440 MWe). Generally, water chemistry at East European VVER plants (about 40 VVER-440 and VVER-1000 units in Ukraine, Bulgaria, Slovakia, Czech Republic, Hungary, Finland and Armenia) is similar to water chemistry at Russian VVER plants. Due to similar design and structural materials some water chemistry improvements were introduced at East European plants after they has been successfully implemented at Russian plants and vice versa. Some water chemistry improvements will be implemented at modern VVER plants under construction in Ukraine, Slovakia, Czech Republic, Iran, China, India. (R.P.)

  3. 42 CFR 1000.10 - General definitions.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false General definitions. 1000.10 Section 1000.10 Public Health OFFICE OF INSPECTOR GENERAL-HEALTH CARE, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS INTRODUCTION; GENERAL DEFINITIONS Definitions § 1000.10 General definitions. In this chapter...

  4. 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...

  5. 7 CFR 1000.6 - Supply plant.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true Supply plant. 1000.6 Section 1000.6 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing Agreements... Definitions § 1000.6 Supply plant. Supply plant means a plant approved by a duly constituted regulatory agency...

  6. 12 CFR 611.1000 - General authority.

    Science.gov (United States)

    2010-01-01

    ... 12 Banks and Banking 6 2010-01-01 2010-01-01 false General authority. 611.1000 Section 611.1000 Banks and Banking FARM CREDIT ADMINISTRATION FARM CREDIT SYSTEM ORGANIZATION Bank Mergers, Consolidations and Charter Amendments § 611.1000 General authority. (a) An amendment to a bank charter may relate...

  7. AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-10-15

    Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

  8. Concepts of radial and angular kinetic energies

    DEFF Research Database (Denmark)

    Dahl, Jens Peder; Schleich, W.P.

    2002-01-01

    We consider a general central-field system in D dimensions and show that the division of the kinetic energy into radial and angular parts proceeds differently in the wave-function picture and the Weyl-Wigner phase-space picture, Thus, the radial and angular kinetic energies are different quantities...

  9. 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.

  10. AP1000, a nuclear central of advanced design; AP1000, una central nuclear de diseno avanzado

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Viais J, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: nhm@nuclear.inin.mx

    2005-07-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  11. Radial gradient and radial deviation radiomic features from pre-surgical CT scans are associated with survival among lung adenocarcinoma patients.

    Science.gov (United States)

    Tunali, Ilke; Stringfield, Olya; Guvenis, Albert; Wang, Hua; Liu, Ying; Balagurunathan, Yoganand; Lambin, Philippe; Gillies, Robert J; Schabath, Matthew B

    2017-11-10

    The goal of this study was to extract features from radial deviation and radial gradient maps which were derived from thoracic CT scans of patients diagnosed with lung adenocarcinoma and assess whether these features are associated with overall survival. We used two independent cohorts from different institutions for training (n= 61) and test (n= 47) and focused our analyses on features that were non-redundant and highly reproducible. To reduce the number of features and covariates into a single parsimonious model, a backward elimination approach was applied. Out of 48 features that were extracted, 31 were eliminated because they were not reproducible or were redundant. We considered 17 features for statistical analysis and identified a final model containing the two most highly informative features that were associated with lung cancer survival. One of the two features, radial deviation outside-border separation standard deviation, was replicated in a test cohort exhibiting a statistically significant association with lung cancer survival (multivariable hazard ratio = 0.40; 95% confidence interval 0.17-0.97). Additionally, we explored the biological underpinnings of these features and found radial gradient and radial deviation image features were significantly associated with semantic radiological features.

  12. Illumination Profile & Dispersion Variation Effects on Radial Velocity Measurements

    Science.gov (United States)

    Grieves, Nolan; Ge, Jian; Thomas, Neil B.; Ma, Bo; Li, Rui; SDSS-III

    2015-01-01

    The Multi-object APO Radial-Velocity Exoplanet Large-Area Survey (MARVELS) measures radial velocities using a fiber-fed dispersed fixed-delay interferometer (DFDI) with a moderate dispersion spectrograph. This setup allows a unique insight into the 2D illumination profile from the fiber on to the dispersion grating. Illumination profile investigations show large changes in the profile over time and fiber location. These profile changes are correlated with dispersion changes and long-term radial velocity offsets, a major problem within the MARVELS radial velocity data. Characterizing illumination profiles creates a method to both detect and correct radial velocity offsets, allowing for better planet detection. Here we report our early results from this study including improvement of radial velocity data points from detected giant planet candidates. We also report an illumination profile experiment conducted at the Kitt Peak National Observatory using the EXPERT instrument, which has a DFDI mode similar to MARVELS. Using profile controlling octagonal-shaped fibers, long term offsets over a 3 month time period were reduced from ~50 m/s to within the photon limit of ~4 m/s.

  13. 21 CFR 589.1000 - Gentian violet.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 6 2010-04-01 2010-04-01 false Gentian violet. 589.1000 Section 589.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) ANIMAL DRUGS... Substances Prohibited From Use in Animal Food or Feed § 589.1000 Gentian violet. The Food and Drug...

  14. Radial transfer effects for poloidal rotation

    Science.gov (United States)

    Hallatschek, Klaus

    2010-11-01

    Radial transfer of energy or momentum is the principal agent responsible for radial structures of Geodesic Acoustic Modes (GAMs) or stationary Zonal Flows (ZF) generated by the turbulence. For the GAM, following a physical approach, it is possible to find useful expressions for the individual components of the Poynting flux or radial group velocity allowing predictions where a mathematical full analysis is unfeasible. Striking differences between up-down symmetric flux surfaces and asymmetric ones have been found. For divertor geometries, e.g., the direction of the propagation depends on the sign of the ion grad-B drift with respect to the X-point, reminiscent of a sensitive determinant of the H-mode threshold. In nonlocal turbulence computations it becomes obvious that the linear energy transfer terms can be completely overwhelmed by the action of the turbulence. In contrast, stationary ZFs are governed by the turbulent radial transfer of momentum. For sufficiently large systems, the Reynolds stress becomes a deterministic functional of the flows, which can be empirically determined from the stress response in computational turbulence studies. The functional allows predictions even on flow/turbulence states not readily obtainable from small amplitude noise, such as certain transport bifurcations or meta-stable states.

  15. United States National Grid for New Mexico, UTM 12, (1000m X 1000m polygons )

    Data.gov (United States)

    Earth Data Analysis Center, University of New Mexico — This is a polygon feature data layer of United States National Grid (1000m x 1000m polygons ) constructed by the Center for Interdisciplinary Geospatial Information...

  16. United States National Grid for New Mexico, UTM 13, (1000m X 1000m polygons )

    Data.gov (United States)

    Earth Data Analysis Center, University of New Mexico — This is a polygon feature data layer of United States National Grid (1000m x 1000m polygons ) constructed by the Center for Interdisciplinary Geospatial Information...

  17. 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

  18. Anterior transposition of the radial nerve--a cadaveric study.

    Science.gov (United States)

    Yakkanti, Madhusudhan R; Roberts, Craig S; Murphy, Joshua; Acland, Robert D

    2008-01-01

    The radial nerve is at risk during the posterior plating of the humerus. The purpose of this anatomic study was to assess the extent of radial nerve dissection required for anterior transposition through the fracture site (transfracture anterior transposition). A cadaver study was conducted approaching the humerus by a posterior midline incision. The extent of dissection of the nerve necessary for plate fixation of the humerus fracture was measured. An osteotomy was created to model a humeral shaft fracture at the spiral groove (OTA classification 12-A2, 12-A3). The radial nerve was then transposed anterior to the humeral shaft through the fracture site. The additional dissection of the radial nerve and the extent of release of soft tissue from the humerus shaft to achieve the transposition were measured. Plating required a dissection of the radial nerve 1.78 cm proximal and 2.13 cm distal to the spiral groove. Transfracture anterior transposition of the radial nerve required an average dissection of 2.24 cm proximal and 2.68 cm distal to the spiral groove. The lateral intermuscular septum had to be released for 2.21 cm on the distal fragment to maintain laxity of the transposed nerve. Transfracture anterior transposition of the radial nerve before plating is feasible with dissection proximal and distal to the spiral groove and elevation of the lateral intermuscular septum. Potential clinical advantages of this technique include enhanced fracture site visualization, application of broader plates, and protection of the radial nerve during the internal fixation.

  19. A New Filtering Algorithm Utilizing Radial Velocity Measurement

    Institute of Scientific and Technical Information of China (English)

    LIU Yan-feng; DU Zi-cheng; PAN Quan

    2005-01-01

    Pulse Doppler radar measurements consist of range, azimuth, elevation and radial velocity. Most of the radar tracking algorithms in engineering only utilize position measurement. The extended Kalman filter with radial velocity measureneut is presented, then a new filtering algorithm utilizing radial velocity measurement is proposed to improve tracking results and the theoretical analysis is also given. Simulation results of the new algorithm, converted measurement Kalman filter, extended Kalman filter are compared. The effectiveness of the new algorithm is verified by simulation results.

  20. KECK NIRSPEC RADIAL VELOCITY OBSERVATIONS OF LATE-M DWARFS

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, Angelle; White, Russel [Department of Astronomy, Georgia State University, One Park Place, Atlanta, GA 30303 (United States); Bailey, John [Department of Astronomy, University of Michigan, 830 Dennison Building, 500 Church Street, Ann Arbor, MI 48109-1042 (United States); Blake, Cullen [Department of Astrophysical Sciences, Princeton University, Peyton Hall, Ivy Lane, Princeton, NJ 08544 (United States); Blake, Geoffrey [Division of Geological and Planetary Sciences, California Institute of Technology, Pasadena, CA 91125 (United States); Cruz, Kelle [Department of Physics and Astronomy, Hunter College, 695 Park Avenue, New York, NY 10065 (United States); Burgasser, Adam J. [Center for Astrophysics and Space Science, University of California San Diego, La Jolla, CA 92093 (United States); Kraus, Adam [Institute for Astronomy, University of Hawaii, 2680 Woodlawn Drive, Honolulu, HI 96822 (United States)

    2012-11-15

    We present the results of an infrared spectroscopic survey of 23 late-M dwarfs with the NIRSPEC echelle spectrometer on the Keck II telescope. Using telluric lines for wavelength calibration, we are able to achieve measurement precisions of down to 45 m s{sup -1} for our late-M dwarfs over a one- to four-year long baseline. Our sample contains two stars with radial velocity (RV) variations of >1000 m s{sup -1}. While we require more measurements to determine whether these RV variations are due to unseen planetary or stellar companions or are the result of starspots known to plague the surface of M dwarfs, we can place upper limits of <40 M{sub J} sin i on the masses of any companions around those two M dwarfs with RV variations of <160 m s{sup -1} at orbital periods of 10-100 days. We have also measured the rotational velocities for all the stars in our late-M dwarf sample and offer our multi-order, high-resolution spectra over 2.0-2.4 {mu}m to the atmospheric modeling community to better understand the atmospheres of late-M dwarfs.

  1. KECK NIRSPEC RADIAL VELOCITY OBSERVATIONS OF LATE-M DWARFS

    International Nuclear Information System (INIS)

    Tanner, Angelle; White, Russel; Bailey, John; Blake, Cullen; Blake, Geoffrey; Cruz, Kelle; Burgasser, Adam J.; Kraus, Adam

    2012-01-01

    We present the results of an infrared spectroscopic survey of 23 late-M dwarfs with the NIRSPEC echelle spectrometer on the Keck II telescope. Using telluric lines for wavelength calibration, we are able to achieve measurement precisions of down to 45 m s –1 for our late-M dwarfs over a one- to four-year long baseline. Our sample contains two stars with radial velocity (RV) variations of >1000 m s –1 . While we require more measurements to determine whether these RV variations are due to unseen planetary or stellar companions or are the result of starspots known to plague the surface of M dwarfs, we can place upper limits of J sin i on the masses of any companions around those two M dwarfs with RV variations of –1 at orbital periods of 10-100 days. We have also measured the rotational velocities for all the stars in our late-M dwarf sample and offer our multi-order, high-resolution spectra over 2.0-2.4 μm to the atmospheric modeling community to better understand the atmospheres of late-M dwarfs.

  2. Photoelectric Radial Velocities, Paper XIX Additional Spectroscopic ...

    Indian Academy of Sciences (India)

    ian velocity curve that does justice to the measurements, but it cannot be expected to have much predictive power. Key words. Stars: late-type—stars: radial velocities—spectroscopic binaries—orbits. 0. Preamble. The 'Redman K stars' are a lot of seventh-magnitude K stars whose radial velocities were first observed by ...

  3. Measurement of the axial and radial diffusivities of a 2D composite material between 500 deg. C and 1500 deg. C; Mesure des diffusivites axiale et radiale d`un composite 2D entre 500 deg. C et 1500 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D.; Beauchene, P.; Casulleras, R.; Bejet, M. [ONERA, 92 - Chatillon (France); Maillet, D.; Sanson, O. [Lemta (France)

    1996-12-31

    A new experimental method of simultaneous measurement of thermal diffusivity along the two main directions of thin composite materials with a ceramic-based matrix has been developed by the ONERA, the French national office of aerospace studies and research. The principle of this method, derived from the `flash` method consists in the heterogeneous insolation of one face of a cylindrical sample (central spot or ring) in order to analyze the thermal transfers along the axial and radial directions of the sample. Experimental development are in progress and will be integrated to a flash diffusion-meter in operation at the ONERA. (J.S.) 11 refs.

  4. Linear theory radial and nonradial pulsations of DA dwarf stars

    International Nuclear Information System (INIS)

    Starrfield, S.; Cox, A.N.; Hodson, S.; Pesnell, W.D.

    1982-01-01

    The Los Alamos stellar envelope and radial linear non-adiabatic computer code, along with a new Los Alamos non-radial code are used to investigate the total hydrogen mass necessary to produce the non-radial instability of DA dwarfs

  5. 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

  6. Westinghouse AP 1000 program status

    International Nuclear Information System (INIS)

    Doehnert, B.

    2002-01-01

    The project 1000 is presented and features are discussed in the paper. Design maturity is characterized by 1300 man-year / $400 million design and testing effort, more than 12 000 design documents completed; 3D computer model developed. It includes structures, equipment, small / large pipe, cable trays, ducts etc. Licensing Maturity is determined by a very thorough and complete NRC review of AP600; 110 man-year effort (NRC) over 6 years, $30 million; independent, confirmatory plant analysis; independent, confirmatory plant testing (ROSA, OSU); over 7400 questions answered, no open items; over 380 meeting with NRC, 43 meetings with ACRS. NRC Design Certification is issued in December 1999. Reasons for developing AP 1000 and design changes are presented. Economic analysis shows an expectation for payback within 20 years. AP1000 provides 75% power uprate for 15% increment in capital cost. AP1000 meets new plant economic targets in the near term

  7. 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.

  8. 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.

  9. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Paranjpe, S.R.; Bhoje, S.B.

    1991-01-01

    Production of electricity during April 1990 - March 1991 was 200 TWh with an increase of 7% over last year. Contribution from coal based thermal is 70%, nuclear 2.5% and 27.5 from hydro. Electricity demand is increasing more than the production growth rate. The programme of installation of 10,000 MWe nuclear capacity in PHWRs by the year 2000 is in progress. 8 x 235 MWe PHWRs are under commissioning and construction. The Government has sanctioned construction of the first 2 x 500 MWe PHWR. Progress in the construction of NPPs is somewhat slow due to industrial infrastructure and financial constraints. There is no public opposition to nuclear power. An intergovernment agreement has been singed between India and the USSR for construction of 2 x 1000 MWe PWRs. FBTR is being operated intermittently up to a power level of 1 MWt without steam generators. Power operation is delayed due to commissioning of a hydrogen leak detection system for the steam generators. (author)

  10. Effect of the radial electric field on turbulence

    International Nuclear Information System (INIS)

    Carreras, B.A.; Lynch, V.E.

    1990-01-01

    For many years, the neoclassical transport theory for three- dimensional magnetic configurations, such as magnetic mirrors, ELMO Bumpy Tori (EBTs), and stellarators, has recognized the critical role of the radial electric field in the confinement. It was in these confinement devices that the first experimental measurements of the radial electric field were made and correlated with confinement losses. In tokamaks, the axisymmetry implies that the neoclassical fluxes are ambipolar and, as a consequence, independent of the radial electric field. However, axisymmetry is not strict in a tokamak with turbulent fluctuations, and near the limiter ambipolarity clearly breaks down. Therefore, the question of the effect of the radial electric field on tokamak confinement has been raised in recent years. In particular, the radial electric field has been proposed to explain the transition from L-mode to H-mode confinement. There is some initial experimental evidence supporting this type of explanation, although there is not yet a self-consistent theory explaining the generation of the electric field and its effect on the transport. Here, a brief review of recent results is presented. 27 refs., 4 figs

  11. A user's evaluation of radial flow HEPA filters

    International Nuclear Information System (INIS)

    Purcell, J.A.

    1992-07-01

    High efficiency particulate air (HEPA) filters of rectangular cross section have been used to remove particulates and the associated radioactivity from air ventilation streams since the advent of nuclear materials processing. Use of round axial flow HEPA filters is also longstanding. The advantages of radial flow filters in a circular configuration have been well demonstrated in UKAEA during the last 5--7 years. An evaluation of radial flow filters for fissile process gloveboxes reveals several substantial benefits in addition to the advantages claimed in UKAEA Facilities. The radial flow filter may be provided in a favorable geometry resulting in improved criticality safety. The filter configuration lends to in-place testing at the glovebox to exhaust duct interface. This will achieve compliance with DOE Order 6430.1A, Section 99.0.2. Preliminary testing at SRS for radial flow filters manufactured by Flanders Filters, Inc. revealed compliance in all the usual specifications for filtration efficiency, pressure differential and materials of construction. An evaluation, further detailed in this report, indicates that the radial flow HEPA filter should be considered for inclusion in new ventilation system designs

  12. Radial extension of drift waves in presence of velocity profiles

    International Nuclear Information System (INIS)

    Sen, S.; Weiland, J.

    1994-01-01

    The effect of a radially varying poloidal velocity field on the recently found radially extended toroidal drift waves is investigated analytically. The role of velocity curvature (υ φ '') is found to have robust effects on the radial model structure of the mode. For a positive value of the curvature (Usually found in the H-mode edges) the radial model envelope, similar to the sheared slab case, becomes fully outgoing. The mode is therefore stable. On the other hand, for a negative value of the curvature (usually observed in the L-mode edges) all the characteristics of conventional drift waves return back. The radial mode envelope reduces to a localized Gaussian shape and the mode is therefore unstable again for typical (magnetic) shear values in tokamaks. Velocity shear (υ φ ??) on the other hand is found to have rather insignificant role both in determining the radial model structure and stability

  13. Parametric cost analysis of a HYLIFE-II power plant

    International Nuclear Information System (INIS)

    Bieri, R.L.

    1991-01-01

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code was adapted to model a power plant using a HYLIFE-2 reactor chamber. The code was then used to examine the dependence of the plant capital costs and the busbar cost of electricity (COE) on a variety of design parameters (type of driver, chamber repetition rate, and net electric power). The results show the most attractive operating space for each set of driver/target assumptions and quantify the benefits of improvements in key design parameters. The base-case plant was a 1000-MW(e) plant containing a reactor vessel driven by an induction linac heavy-ion accelerator, run at 8 Hz with a driver energy of 6.73 MJ and a target yield of 350 MJ. The total direct cost for this plant was $2.6 billion. (All costs in this paper are given in equivalent 1988 dollars.) The COE was 8.5 cents/(kWh). The COE and total capital costs for a 1000-MW(e) base plant are nearly independent of the chosen combination of repetition rate and driver energy for a driver operating between 4 and 10 Hz. For comparison, the COE for a coal or future fission plant would be 4.5--5.5 cents/(kWh). The COE for a 1000-MW(e) plant could be reduced to 7.5 cents/(kWh) by using advanced targets and could be cut to 6.5 cents/(kWh) with conventional targets, if the driver cost could be cut in half. There is a large economy of scale with heavy-ion-driven inertial confinement fusion (ICF) plants. A 2000-MW(e) plant with a heavy-ion driver and a HYLIFE-2 chamber would have a COE of only 5.8 cents/(kWh)

  14. AP1000{sup TM} plant modularization

    Energy Technology Data Exchange (ETDEWEB)

    Cantarero L, C.; Demetri, K. J. [Westinghouse Electric Co., 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Quintero C, F. P., E-mail: cantarc@westinghouse.com [Westinghouse Electric Spain, Padilla 17, 28006 Madrid (Spain)

    2016-09-15

    The AP1000{sup TM} plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

  15. AP1000"T"M plant modularization

    International Nuclear Information System (INIS)

    Cantarero L, C.; Demetri, K. J.; Quintero C, F. P.

    2016-09-01

    The AP1000"T"M plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

  16. Combined Radial and Femoral Access Strategy and Radial-Femoral Rendezvous in Patients With Long and Complex Iliac Occlusions.

    Science.gov (United States)

    Hanna, Elias B; Mogabgab, Owen N; Baydoun, Hassan

    2018-01-01

    We present cases of complex, calcified iliac occlusive disease revascularized via a combined radial-femoral access strategy. Through a 6-French, 125-cm transradial guiding catheter, antegrade guidewires and catheters are advanced into the iliac occlusion, while retrograde devices are advanced transfemorally. The transradial and transfemoral channels communicate, allowing the devices to cross the occlusion into the true lumen (radial-femoral antegrade-retrograde rendezvous).

  17. A novel structure of permanent-magnet-biased radial hybrid magnetic bearing

    International Nuclear Information System (INIS)

    Sun Jinji; Fang Jiancheng

    2011-01-01

    The paper proposes a novel structure for a permanent-magnet-biased radial hybrid magnetic bearing. Based on the air gap between the rotor and stator of traditional radial hybrid magnetic bearings, a subsidiary air gap is first constructed between the permanent magnets and the inner magnetic parts. Radial magnetic bearing makes X and Y magnetic fields independent of each other with separate stator poles, and the subsidiary air gap makes control flux to a close loop. As a result, magnetic field coupling of the X and Y channels is decreased significantly by the radial hybrid magnetic bearing and makes it easier to design control systems. Then an external rotor structure is designed into the radial hybrid magnetic bearing. The working principle of the radial hybrid magnetic bearing and its mathematical model is discussed. Finally, a non-linear magnetic network method is proposed to analyze the radial hybrid magnetic bearing. Simulation results indicate that magnetic fields in the two channels of the proposed radial hybrid magnetic bearing decouple well from each other.

  18. A novel structure of permanent-magnet-biased radial hybrid magnetic bearing

    Energy Technology Data Exchange (ETDEWEB)

    Sun Jinji, E-mail: sunjinji@aspe.buaa.edu.c [Key Laboratory of Fundamental Science for National Defense, Novel Inertial Instrument and Navigation System Technology, School of Instrument Science and Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, 100191 (China); Fang Jiancheng [Key Laboratory of Fundamental Science for National Defense, Novel Inertial Instrument and Navigation System Technology, School of Instrument Science and Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, 100191 (China)

    2011-01-15

    The paper proposes a novel structure for a permanent-magnet-biased radial hybrid magnetic bearing. Based on the air gap between the rotor and stator of traditional radial hybrid magnetic bearings, a subsidiary air gap is first constructed between the permanent magnets and the inner magnetic parts. Radial magnetic bearing makes X and Y magnetic fields independent of each other with separate stator poles, and the subsidiary air gap makes control flux to a close loop. As a result, magnetic field coupling of the X and Y channels is decreased significantly by the radial hybrid magnetic bearing and makes it easier to design control systems. Then an external rotor structure is designed into the radial hybrid magnetic bearing. The working principle of the radial hybrid magnetic bearing and its mathematical model is discussed. Finally, a non-linear magnetic network method is proposed to analyze the radial hybrid magnetic bearing. Simulation results indicate that magnetic fields in the two channels of the proposed radial hybrid magnetic bearing decouple well from each other.

  19. Detonation in supersonic radial outflow

    KAUST Repository

    Kasimov, Aslan R.

    2014-11-07

    We report on the structure and dynamics of gaseous detonation stabilized in a supersonic flow emanating radially from a central source. The steady-state solutions are computed and their range of existence is investigated. Two-dimensional simulations are carried out in order to explore the stability of the steady-state solutions. It is found that both collapsing and expanding two-dimensional cellular detonations exist. The latter can be stabilized by putting several rigid obstacles in the flow downstream of the steady-state sonic locus. The problem of initiation of standing detonation stabilized in the radial flow is also investigated numerically. © 2014 Cambridge University Press.

  20. 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.

  1. Generation of live offspring from vitrified embryos with synthetic polymers SuperCool X-1000 and SuperCool Z-1000.

    Science.gov (United States)

    Marco-Jimenez, F; Jimenez-Trigos, E; Lavara, R; Vicente, J S

    2014-01-01

    Ice growth and recrystallisation are considered important factors in determining vitrification outcomes. Synthetic polymers inhibit ice formation during cooling or warming of the vitrification process. The aim of this study was to assess the effect of adding commercially available synthetic polymers SuperCool X-1000 and SuperCool Z-1000 to vitrification media on in vivo development competence of rabbit embryos. Four hundred and thirty morphologically normal embryos recovered at 72 h of gestation were used. The vitrification media contained 20% dimethyl sulphoxide and 20% ethylene glycol, either alone or in combination with 1% of SuperCool X-1000 and 1% SuperCool. Our results show that embryos can be successfully vitrified using SuperCool X-1000 and SuperCool Z-1000 and when embryos are transferred, live offspring can be successfully produced. In conclusion, our results demonstrated that we succeeded for the first time in obtaining live offspring after vitrification of embryos using SuperCool X-1000 and SuperCool Z-1000 polymers.

  2. Statistical characterization of Earth’s heterogeneities from seismic scattering

    Science.gov (United States)

    Zheng, Y.; Wu, R.

    2009-12-01

    The distortion of a teleseismic wavefront carries information about the heterogeneities through which the wave propagates and it is manifestited as logarithmic amplitude (logA) and phase fluctuations of the direct P wave recorded by a seismic network. By cross correlating the fluctuations (e.g., logA-logA or phase-phase), we obtain coherence functions, which depend on spatial lags between stations and incident angles between the incident waves. We have mathematically related the depth-dependent heterogeneity spectrum to the observable coherence functions using seismic scattering theory. We will show that our method has sharp depth resolution. Using the HiNet seismic network data in Japan, we have inverted power spectra for two depth ranges, ~0-120km and below ~120km depth. The coherence functions formed by different groups of stations or by different groups of earthquakes at different back azimuths are similar. This demonstrates that the method is statistically stable and the inhomogeneities are statistically stationary. In both depth intervals, the trend of the spectral amplitude decays from large scale to small scale in a power-law fashion with exceptions at ~50km for the logA data. Due to the spatial spacing of the seismometers, only information from length scale 15km to 200km is inverted. However our scattering method provides new information on small to intermediate scales that are comparable to scales of the recycled materials and thus is complimentary to the global seismic tomography which reveals mainly large-scale heterogeneities on the order of ~1000km. The small-scale heterogeneities revealed here are not likely of pure thermal origin. Therefore, the length scale and strength of heterogeneities as a function of depth may provide important constraints in mechanical mixing of various components in the mantle convection.

  3. Experimental stress analysis of the attachment region of hemispherical shells with attached nozzles. Part 2b. Radial nozzle 7.875 in. O.D.--7.500 in. I.D. 10.00 in. penetration

    International Nuclear Information System (INIS)

    Maxwell, R.L.; Holland, R.W.; Stengl, G.R.

    1970-06-01

    The report presents the results of investigations conducted on a hemisphere with a radial nozzle of 7.875'' O.D. and 7.500'' I.D. and 10'' penetration into the hemisphere. Stress values were determined for the following five types of loadings: (1) internal pressure applied to the hemisphere and nozzle assembly, (2) an axial load applied collinear with nozzle, (3) a pure bending moment, or axial couple, applied to the nozzle, (4) a transverse or shear load applied normal to the nozzle, and (5) a pure torque applied in the radial plane of the nozzle

  4. Radial nerve dysfunction (image)

    Science.gov (United States)

    The radial nerve travels down the arm and supplies movement to the triceps muscle at the back of the upper arm. ... the wrist and hand. The usual causes of nerve dysfunction are direct trauma, prolonged pressure on the ...

  5. Variable stator radial turbine

    Science.gov (United States)

    Rogo, C.; Hajek, T.; Chen, A. G.

    1984-01-01

    A radial turbine stage with a variable area nozzle was investigated. A high work capacity turbine design with a known high performance base was modified to accept a fixed vane stagger angle moveable sidewall nozzle. The nozzle area was varied by moving the forward and rearward sidewalls. Diffusing and accelerating rotor inlet ramps were evaluated in combinations with hub and shroud rotor exit rings. Performance of contoured sidewalls and the location of the sidewall split line with respect to the rotor inlet was compared to the baseline. Performance and rotor exit survey data are presented for 31 different geometries. Detail survey data at the nozzle exit are given in contour plot format for five configurations. A data base is provided for a variable geometry concept that is a viable alternative to the more common pivoted vane variable geometry radial turbine.

  6. Radial-probe EBUS for the diagnosis of peripheral pulmonary lesions

    Directory of Open Access Journals (Sweden)

    Marcia Jacomelli

    Full Text Available ABSTRACT Objective: Conventional bronchoscopy has a low diagnostic yield for peripheral pulmonary lesions. Radial-probe EBUS employs a rotating ultrasound transducer at the end of a probe that is passed through the working channel of the bronchoscope. Radial-probe EBUS facilitates the localization of peripheral pulmonary nodules, thus increasing the diagnostic yield. The objective of this study was to present our initial experience using radial-probe EBUS in the diagnosis of peripheral pulmonary lesions at a tertiary hospital. Methods: We conducted a retrospective analysis of 54 patients who underwent radial-probe EBUS-guided bronchoscopy for the investigation of pulmonary nodules or masses between February of 2012 and September of 2013. Radial-probe EBUS was performed with a flexible 20-MHz probe, which was passed through the working channel of the bronchoscope and advanced through the bronchus to the target lesion. For localization of the lesion and for collection procedures (bronchial brushing, transbronchial needle aspiration, and transbronchial biopsy, we used fluoroscopy. Results: Radial-probe EBUS identified 39 nodules (mean diameter, 1.9 ± 0.7 cm and 19 masses (mean diameter, 4.1 ± 0.9 cm. The overall sensitivity of the method was 66.7% (79.5% and 25.0%, respectively, for lesions that were visible and not visible by radial-probe EBUS. Among the lesions that were visible by radial-probe EBUS, the sensitivity was 91.7% for masses and 74.1% for nodules. The complications were pneumothorax (in 3.7% and bronchial bleeding, which was controlled bronchoscopically (in 9.3%. Conclusions: Radial-probe EBUS shows a good safety profile, a low complication rate, and high sensitivity for the diagnosis of peripheral pulmonary lesions.

  7. Sharp Dissection versus Electrocautery for Radial Artery Harvesting

    Science.gov (United States)

    Marzban, Mehrab; Arya, Reza; Mandegar, Mohammad Hossein; Karimi, Abbas Ali; Abbasi, Kiomars; Movahed, Namvar; Abbasi, Seyed Hesameddin

    2006-01-01

    Radial arteries have been increasingly used during the last decade as conduits for coronary artery revascularization. Although various harvesting techniques have been described, there has been little comparative study of arterial damage and patency. A radial artery graft was used in 44 consecutive patients, who were randomly divided into 2 groups. In the 1st group, the radial artery was harvested by sharp dissection and in the 2nd, by electrocautery. These groups were compared with regard to radial artery free flow, harvest time, number of clips used, complications, and endothelial damage. Radial artery free flow before and after intraluminal administration of papaverine was significantly greater in the electrocautery group (84.3 ± 50.7 mL/min and 109.7 ± 68.5 mL/min) than in the sharp-dissection group (52.9 ± 18.3 mL/min and 69.6 ± 28.2 mL/ min) (P =0.003). Harvesting time by electrocautery was significantly shorter (25.4 ± 4.3 min vs 34.4 ± 5.9 min) (P =0.0001). Electrocautery consumed an average of 9.76 clips, versus 22.45 clips consumed by sharp dissection. The 2 groups were not different regarding postoperative complications, except for 3 cases of temporary paresthesia of the thumb in the electrocautery group; histopathologic examination found no endothelial damage. We conclude that radial artery harvesting by electrocautery is faster and more economical than harvesting by sharp dissection and is associated with better intraoperative flow and good preservation of endothelial integrity. PMID:16572861

  8. CHARACTERIZING LANDSCAPE SPATIAL HETEROGENEITY USING SEMIVARIOGRAM PARAMETERS DERIVED FROM NDVI IMAGES

    Directory of Open Access Journals (Sweden)

    Eduarda Martiniano de Oliveira Silveira

    2017-12-01

    Full Text Available Assuming a relationship between landscape heterogeneity and measures of spatial dependence by using remotely sensed data, the aim of this work was to evaluate the potential of semivariogram parameters, derived from satellite images with different spatial resolutions, to characterize landscape spatial heterogeneity of forested and human modified areas. The NDVI (Normalized Difference Vegetation Index was generated in an area of Brazilian amazon tropical forest (1,000 km².We selected samples (1 x 1 km from forested and human modified areas distributed throughout the study area, to generate the semivariogram and extract the sill (σ²-overall spatial variability of the surface property and range (φ-the length scale of the spatial structures of objects parameters. The analysis revealed that image spatial resolution influenced the sill and range parameters. The average sill and range values increase from forested to human modified areas and the greatest between-class variation was found for LANDSAT 8 imagery, indicating that this image spatial resolution is the most appropriate for deriving sill and range parameters with the intention of describing landscape spatial heterogeneity. By combining remote sensing and geostatistical techniques, we have shown that the sill and range parameters of semivariograms derived from NDVI images are a simple indicator of landscape heterogeneity and can be used to provide landscape heterogeneity maps to enable researchers to design appropriate sampling regimes. In the future, more applications combining remote sensing and geostatistical features should be further investigated and developed, such as change detection and image classification using object-based image analysis (OBIA approaches.

  9. Radial distribution of ions in pores with a surface charge

    NARCIS (Netherlands)

    Stegen, J.H.G. van der; Görtzen, J.; Kuipers, J.A.M.; Hogendoorn, J.A.; Versteeg, G.F.

    2001-01-01

    A sorption model applicable to calculate the radial equilibrium concentrations of ions in the pores of ion-selective membranes with a pore structure is developed. The model is called the radial uptake model. Because the model is applied to a Nafion sulfonic layer with very small pores and the radial

  10. The Matlab Radial Basis Function Toolbox

    Directory of Open Access Journals (Sweden)

    Scott A. Sarra

    2017-03-01

    Full Text Available Radial Basis Function (RBF methods are important tools for scattered data interpolation and for the solution of Partial Differential Equations in complexly shaped domains. The most straight forward approach used to evaluate the methods involves solving a linear system which is typically poorly conditioned. The Matlab Radial Basis Function toolbox features a regularization method for the ill-conditioned system, extended precision floating point arithmetic, and symmetry exploitation for the purpose of reducing flop counts of the associated numerical linear algebra algorithms.

  11. Stellar Angular Momentum Distributions and Preferential Radial Migration

    Science.gov (United States)

    Wyse, Rosemary; Daniel, Kathryne J.

    2018-04-01

    I will present some results from our recent investigations into the efficiency of radial migration in stellar disks of differing angular momentum distributions, within a given adopted 2D spiral disk potential. We apply to our models an analytic criterion that determines whether or not individual stars are in orbits that could lead to radial migration around the corotation resonance. We couch our results in terms of the local stellar velocity dispersion and find that the fraction of stars that could migrate radially decreases as the velocity dispersion increases. I will discuss implications and comparisons with the results of other approaches.

  12. Research on Radial Vibration of a Circular Plate

    Directory of Open Access Journals (Sweden)

    Wei Liu

    2016-01-01

    Full Text Available Radial vibration of the circular plate is presented using wave propagation approach and classical method containing Bessel solution and Hankel solution for calculating the natural frequency theoretically. In cylindrical coordinate system, in order to obtain natural frequency, propagation and reflection matrices are deduced at the boundaries of free-free, fixed-fixed, and fixed-free using wave propagation approach. Furthermore, radial phononic crystal is constructed by connecting two materials periodically for the analysis of band phenomenon. Also, Finite Element Simulation (FEM is adopted to verify the theoretical results. Finally, the radial and piezoelectric effects on the band are also discussed.

  13. 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)

  14. 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.

  15. Radial-probe EBUS for the diagnosis of peripheral pulmonary lesions.

    Science.gov (United States)

    Jacomelli, Marcia; Demarzo, Sergio Eduardo; Cardoso, Paulo Francisco Guerreiro; Palomino, Addy Lidvina Mejia; Figueiredo, Viviane Rossi

    2016-01-01

    Conventional bronchoscopy has a low diagnostic yield for peripheral pulmonary lesions. Radial-probe EBUS employs a rotating ultrasound transducer at the end of a probe that is passed through the working channel of the bronchoscope. Radial-probe EBUS facilitates the localization of peripheral pulmonary nodules, thus increasing the diagnostic yield. The objective of this study was to present our initial experience using radial-probe EBUS in the diagnosis of peripheral pulmonary lesions at a tertiary hospital. We conducted a retrospective analysis of 54 patients who underwent radial-probe EBUS-guided bronchoscopy for the investigation of pulmonary nodules or masses between February of 2012 and September of 2013. Radial-probe EBUS was performed with a flexible 20-MHz probe, which was passed through the working channel of the bronchoscope and advanced through the bronchus to the target lesion. For localization of the lesion and for collection procedures (bronchial brushing, transbronchial needle aspiration, and transbronchial biopsy), we used fluoroscopy. Radial-probe EBUS identified 39 nodules (mean diameter, 1.9 ± 0.7 cm) and 19 masses (mean diameter, 4.1 ± 0.9 cm). The overall sensitivity of the method was 66.7% (79.5% and 25.0%, respectively, for lesions that were visible and not visible by radial-probe EBUS). Among the lesions that were visible by radial-probe EBUS, the sensitivity was 91.7% for masses and 74.1% for nodules. The complications were pneumothorax (in 3.7%) and bronchial bleeding, which was controlled bronchoscopically (in 9.3%). Radial-probe EBUS shows a good safety profile, a low complication rate, and high sensitivity for the diagnosis of peripheral pulmonary lesions. A broncoscopia convencional possui baixo rendimento diagnóstico para lesões pulmonares periféricas. A ecobroncoscopia radial (EBUS radial) emprega um transdutor ultrassonográfico rotatório na extremidade de uma sonda que é inserida no canal de trabalho do broncoscópio. O EBUS

  16. Radial vibration and ultrasonic field of a long tubular ultrasonic radiator.

    Science.gov (United States)

    Shuyu, Lin; Zhiqiang, Fu; Xiaoli, Zhang; Yong, Wang; Jing, Hu

    2013-09-01

    The radial vibration of a metal long circular tube is studied analytically and its electro-mechanical equivalent circuit is obtained. Based on the equivalent circuit, the radial resonance frequency equation is derived. The theoretical relationship between the radial resonance frequency and the geometrical dimensions is studied. Finite element method is used to simulate the radial vibration and the radiated ultrasonic field and the results are compared with those from the analytical method. It is concluded that the radial resonance frequency for a solid metal rod is larger than that for a metal tube with the same outer radius. The radial resonance frequencies from the analytical method are in good agreement with those from the numerical method. Based on the acoustic field analysis, it is concluded that the long metal tube with small wall thickness is superior to that with large wall thickness in producing radial vibration and ultrasonic radiation. Therefore, it is expected to be used as an effective radial ultrasonic radiator in ultrasonic sewage treatment, ultrasonic antiscale and descaling and other ultrasonic liquid handling applications. Copyright © 2013 Elsevier B.V. All rights reserved.

  17. Analysis of radionuclide dispersion at normal condition for AEC 1000 MW reactor power

    International Nuclear Information System (INIS)

    Sri Kuntjoro

    2010-01-01

    Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC) 3,568 MWth Power Reactor, equal to the 1,000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90° one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL. computer code program. ORIGEN2 applied to determine radionuclide inventory emerged in the reactor. Hereinafter, by using Emerald Normal Computer code is calculated source term reaching the reactor stack. To analyze dose received by population is done by using PC-CREAM computer code. Calculation done for one and two PLTN attached in site candidate of plants. The result showed is that the highest radionuclide release for one PLTN is at 1 km distance and to 9 th zone toward ( 19.25° ) and for two PLTN is at 1 km distance and to 10 th zone toward (21.75° ). Radionuclide which up to population through two pathways that are foodstuff and inhalation. To foodstuff comes from radionuclide I 131 , and the biggest passed from milk product with 53.40 % for one and also two PLTN For inhalation pathway the highest radionuclide contribution come from Kr 85m is about 53.80 %. The highest total dose received by population is at 1 Km distance received by baby that is 4.10 µSi and 11.26 µSi for one and two PLTN respectively. Those result are very small compared to the maximum permission dose to population issued by regulatory body that is equal to 1 mSi. (author)

  18. 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%

  19. 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.

  20. Radial Fuzzy Systems

    Czech Academy of Sciences Publication Activity Database

    Coufal, David

    2017-01-01

    Roč. 319, 15 July (2017), s. 1-27 ISSN 0165-0114 R&D Projects: GA MŠk(CZ) LD13002 Institutional support: RVO:67985807 Keywords : fuzzy systems * radial functions * coherence Subject RIV: BA - General Mathematics OBOR OECD: Computer sciences, information science, bioinformathics (hardware development to be 2.2, social aspect to be 5.8) Impact factor: 2.718, year: 2016

  1. Acesso radial em intervenções coronarianas percutâneas: panorama atual brasileiro Acceso radial en intervenciones coronarias percutáneas: panorama actual brasileño Radial approach in percutaneous coronary interventions: current status in Brazil

    Directory of Open Access Journals (Sweden)

    Pedro Beraldo de Andrade

    2011-04-01

    Full Text Available FUNDAMENTO: Embora a técnica radial exiba resultados incontestáveis na redução de complicações vasculares e ocorrência de sangramento grave quando comparada à técnica femoral, seu emprego permanece restrito a poucos centros que a elegeram como via de acesso preferencial. OBJETIVO: Avaliar o cenário atual das intervenções coronarianas percutâneas no Brasil quanto à utilização da via de acesso radial. MÉTODOS: Análise dos dados cadastrados de forma espontânea na Central Nacional de Intervenções Cardiovasculares (CENIC durante o quadriênio de 2005-2008, o que totaliza 83.376 procedimentos. RESULTADOS: A técnica radial foi utilizada em 12,6% dos procedimentos efetivados, e a técnica femoral, em 84,3%. Os 3,1% restantes foram representados pela dissecção ou punção braquial. Com uma taxa de sucesso de 97,5%, a opção pelo acesso radial associou-se à redução significativa de complicações vasculares quando comparado ao femoral (2,5% versus 3,6%, p FUNDAMENTO: Aunque la técnica radial exhiba resultados incontestables en la reducción de complicaciones vasculares y ocurrencia de sangrado grave cuando es comparada a la técnica femoral, su empleo permanece restringido a pocos centros que la eligieron como vía de acceso preferencial. OBJETIVO:Evaluar el escenario actual de las intervenciones coronarias percutáneas en el Brasil en cuanto a la utilización de la vía de acceso radial. MÉTODOS:Análisis de los datos registrados de forma espontánea en la Central Nacional de Intervenciones Cardiovasculares (CENIC durante el cuatrienio de 2005-2008, lo que totaliza 83.376 procedimientos. RESULTADOS:La técnica radial fue utilizada en 12,6% de los procedimientos efectuados, y la técnica femoral, en 84,3%. Los 3,1% restantes fueron representados por la disección o punción braquial. Con una tasa de éxito de 97,5%, la opción por el acceso radial se asoció a la reducción significativa de complicaciones vasculares cuando

  2. 7 CFR 1940.1000 - OMB control number.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 13 2010-01-01 2009-01-01 true OMB control number. 1940.1000 Section 1940.1000 Agriculture Regulations of the Department of Agriculture (Continued) RURAL HOUSING SERVICE, RURAL BUSINESS....1000 OMB control number. The collection of information requirements contained in this regulation has...

  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. 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.

  5. 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.

  6. Velocidades radiales en Collinder 121

    Science.gov (United States)

    Arnal, M.; Morrell, N.

    Se han llevado a cabo observaciones espectroscópicas de unas treinta estrellas que son posibles miembros del cúmulo abierto Collinder 121. Las mismas fueron realizadas con el telescopio de 2.15m del Complejo Astronómico El Leoncito (CASLEO). El análisis de las velocidades radiales derivadas del material obtenido, confirma la realidad de Collinder 121, al menos desde el punto de vista cinemático. La velocidad radial baricentral (LSR) del cúmulo es de +17 ± 3 km.s-1. Esta velocidad coincide, dentro de los errores, con la velocidad radial (LSR) de la nebulosa anillo S308, la cual es de ~20 ± 10 km.s-1. Como S308 se encuentra físicamente asociada a la estrella Wolf-Rayet HD~50896, es muy probable que esta última sea un miembro de Collinder 121. Desde un punto de vista cinemático, la supergigante roja HD~50877 (K3Iab) también pertenecería a Collinder 121. Basándonos en la pertenencia de HD~50896 a Collinder 121, y en la interacción encontrada entre el viento de esta estrella y el medio interestelar circundante a la misma, se estima para este cúmulo una distancia del orden de 1 kpc.

  7. 26 CFR 1.998-1.1000 - [Reserved

    Science.gov (United States)

    2010-04-01

    ... 26 Internal Revenue 10 2010-04-01 2010-04-01 false [Reserved] 1.998-1.1000 Section 1.998-1.1000 Internal Revenue INTERNAL REVENUE SERVICE, DEPARTMENT OF THE TREASURY (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES Domestic International Sales Corporations §§ 1.998-1.1000 [Reserved] ...

  8. 7 CFR 4284.1000 - OMB control number.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 15 2010-01-01 2010-01-01 false OMB control number. 4284.1000 Section 4284.1000 Agriculture Regulations of the Department of Agriculture (Continued) RURAL BUSINESS-COOPERATIVE SERVICE AND RURAL UTILITIES SERVICE, DEPARTMENT OF AGRICULTURE GRANTS Value-Added Producer Grants § 4284.1000 OMB...

  9. MCNP Simulations of End Flux Peaking in ACR-1000, 2.4 wt % {sup 235}U Fuel Bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Ian; Donnelly, Jim [Atomic Energy of Canada Limited (AECL), 2251 Speakman Drive, Mississauga, ON, L5K 1B2 (Canada)

    2008-07-01

    This paper examines the end flux peaking in ACR-1000 fuel bundles. Reactor physics simulations are performed with MCNP to assess the steady state end-flux peaking in an infinite lattice of ACR fuel, as well as to quantify the peaking that occurs during refuelling. 3-dimensional MCNP models are created based on the detailed geometry of the fuel bundle. Detailed position-dependent fuel compositions are obtained from MONTEBURNS which couples MCNP and ORIGIN2.2. Axial and radial power profiles are obtained for both fresh and mid-burnup fuel bundles in an infinite lattice. Subsequently an assessment of the impact of a refuelling transient on the power profiles is performed. The refuelling transient is found to increase the end flux peaking in the region adjacent to light water. (authors)

  10. Improving sub-pixel imperviousness change prediction by ensembling heterogeneous non-linear regression models

    Directory of Open Access Journals (Sweden)

    Drzewiecki Wojciech

    2016-12-01

    Full Text Available In this work nine non-linear regression models were compared for sub-pixel impervious surface area mapping from Landsat images. The comparison was done in three study areas both for accuracy of imperviousness coverage evaluation in individual points in time and accuracy of imperviousness change assessment. The performance of individual machine learning algorithms (Cubist, Random Forest, stochastic gradient boosting of regression trees, k-nearest neighbors regression, random k-nearest neighbors regression, Multivariate Adaptive Regression Splines, averaged neural networks, and support vector machines with polynomial and radial kernels was also compared with the performance of heterogeneous model ensembles constructed from the best models trained using particular techniques.

  11. Radial supports of face motors with slack compensation

    Energy Technology Data Exchange (ETDEWEB)

    Kuznetsova, I I; Gelman, A B; Krekina, T V

    1982-01-01

    The design of a radial support of a face motor with slack compensation is described, and gives the results of field tests which confirm the performance capacity of the experimental support both from the viewpoint of durability, and in relation to preventing radial slack of the face motor shaft.

  12. Queratotomía radial versus miniqueratotomía radial: Experiencia en el Hospital "Ramón Pando Ferrer" Radial keratotomy versus radial minikeratotomy: Experience in "Ramón Pando Ferrer" Hospital

    Directory of Open Access Journals (Sweden)

    José Edilberto Pacheco Serrano

    2000-06-01

    Full Text Available La miniqueratotomía radial se viene realizando desde 1995. Se plantea que incisiones más cortas tienen el mismo efecto y producen menos debilidad corneal, ya que disminuye la susceptibilidad a sufrir complicaciones graves provenientes de traumas de la vida cotidiana. Esta idea nos motivó a realizar un estudio para observar el comportamiento de incisiones más cortas en nuestro centro y, en caso de resultados positivos, implementar la técnica de manera que nuestros pacientes puedan beneficiarse de ella. Se comparan resultados de la aplicación de dos técnicas quirúrgicas refractivas para corrección de miopía entre leve y moderada. Se seleccionaron 38 pacientes entre 20 y 40 años de edad, con miopías entre -2 y -6 dioptrías y astigmatismo no mayor a -0,75 dioptrías. Se realizó queratotomía radial convencional en el ojo derecho y miniqueratotomía radial en el ojo izquierdo del mismo paciente. Las variaciones obtenidas en promedio fueron, en el ojo derecho: la esfera (en dioptrías D de -3,38 a -0,32, cilindro de -0,48 a -0,45 D, la queratometría de 44,75 a 41,21 D. En el ojo izquierdo: la esfera de -3,38 D a -0,44 D, cilindro de -0,44 D a -0,38 D, la queratometría de 44,83 a 41,80 D. Hubo una mejoría de la agudeza visual sin cristales de 0,61 en el ojo derecho y 0,59 en el ojo izquierdo. Las dos técnicas no mostraron diferencias estadísticamente significativas, con el beneficio de que la nueva técnica disminuye el riesgo de ruptura postraumática, según la bibliografía revisada, a causa de la menor injuria corneal.In this hospital, radial keratotomy is performed sice 1995. We propose that shorter incisions have some effect and cause less corneal weakness, since dicreases susceptibility to severe complications from traumata of daily life. This notion encouraged us to carry out a study to observe behaviour of shorter incisions in our service, and in the event of positive results, implementation of the technique so that our

  13. Pseudarthrosis of radial shaft with dislocation of heads of radial and ulnar bones (case report

    Directory of Open Access Journals (Sweden)

    M. E. Puseva

    2013-01-01

    Full Text Available The authors presented a rare clinical case - the injury of forearm complicated by the formation of the pseudarthrosis of the radial shaft in combination with old dislocation of heads the radius and ulna. The differentiated approach to the choice of surgical tactics was proposed, which consists of several consistent stages: taking free autotransplant from the crest of iliac bone, resection of pseudarthrosis of radius with replacement of the bone defect by the graft for restoration of anatomic length, conducting combined strained osteosynthesis and elimination of dislocation of a head of radial and ulnar bones by transosseous osteosynthesis. The chosen treatment strategy allowed to restore the anatomy and function of the upper extremity.

  14. 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.

  15. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis

    International Nuclear Information System (INIS)

    Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.

    2004-01-01

    The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)

  16. 7 CFR 1000.14 - Other source milk.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true Other source milk. 1000.14 Section 1000.14 Agriculture... and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Definitions § 1000.14 Other source milk. Other source milk means all skim milk and butterfat contained in or...

  17. Regional energy efficiency, carbon emission performance and technology gaps in China: A meta-frontier non-radial directional distance function analysis

    International Nuclear Information System (INIS)

    Yao, Xin; Zhou, Hongchen; Zhang, Aizhen; Li, Aijun

    2015-01-01

    At present, China is the largest primary energy consumer and carbon emitter in the world. Meantime, China is a large transitional economy with significant regional gaps. Against such backgrounds, the calculated results of energy and carbon performance indicators may be biased, without considering heterogeneity across regions. To this end, after incorporating region-heterogeneity, this paper provides detailed information, regarding energy efficiency, carbon emission performance and the potential of carbon emission reductions from regional perspectives, which may be important and useful for policy makers. Our main findings are as follows. Firstly, there is significant group-heterogeneity across regions in China, in terms of energy efficiency and carbon emission performance. Secondly, there are no considerable differences between total-factor and single-factor performance indices, since there is limited substitutability between energy inputs and other production inputs. Finally, significant carbon emission reductions can be made by “catching up” for regions with low energy efficiency and carbon emission performance. Looking ahead, the Chinese government should adopt measures to promote improvements in terms of energy efficiency and carbon emission performance in the short term. -- Highlights: •We adopt a meta-frontier non-radial directional distance function analysis. •We provide detailed information regarding energy and carbon emission performance. •We find that there is significant region-heterogeneity in China. •There are no large differences between total- and single-factor performance indices. •It can make great contributions to carbon emission reductions by “catching up”

  18. 25 CFR 39.1000 - Purpose and scope.

    Science.gov (United States)

    2010-04-01

    ... 25 Indians 1 2010-04-01 2010-04-01 false Purpose and scope. 39.1000 Section 39.1000 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR EDUCATION THE INDIAN SCHOOL EQUALIZATION PROGRAM Administrative Cost Formula § 39.1000 Purpose and scope. The purpose of this subpart is to provide funds at the...

  19. 27 CFR 19.1000 - Reconsignment in transit.

    Science.gov (United States)

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Reconsignment in transit. 19.1000 Section 19.1000 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE..., Withdrawals and Transfers § 19.1000 Reconsignment in transit. When, prior to or on arrival at the premises of...

  20. 24 CFR 35.1000 - Purpose and applicability.

    Science.gov (United States)

    2010-04-01

    ... 24 Housing and Urban Development 1 2010-04-01 2010-04-01 false Purpose and applicability. 35.1000 Section 35.1000 Housing and Urban Development Office of the Secretary, Department of Housing and Urban..., Support Services, or Operation § 35.1000 Purpose and applicability. (a) The purpose of this subpart K is...

  1. Manufacturing of Precision Forgings by Radial Forging

    International Nuclear Information System (INIS)

    Wallner, S.; Harrer, O.; Buchmayr, B.; Hofer, F.

    2011-01-01

    Radial forging is a multi purpose incremental forging process using four tools on the same plane. It is widely used for the forming of tool steels, super alloys as well as titanium- and refractory metals. The range of application goes from reducing the diameters of shafts, tubes, stepped shafts and axels, as well as for creating internal profiles for tubes in Near-Net-Shape and Net-Shape quality. Based on actual development of a weight optimized transmission input shaft, the specific features of radial forging technology is demonstrated. Also a Finite Element Model for the simulation of the process is shown which leads to reduced pre-processing effort and reduced computing time compared to other published simulation methods for radial forging. The finite element model can be applied to quantify the effects of different forging strategies.

  2. Radially truncated galactic discs

    NARCIS (Netherlands)

    Grijs, R. de; Kregel, M.; Wesson, K H

    2000-01-01

    Abstract: We present the first results of a systematic analysis of radially truncatedexponential discs for four galaxies of a sample of disc-dominated edge-onspiral galaxies. Edge-on galaxies are very useful for the study of truncatedgalactic discs, since we can follow their light distributions out

  3. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)

    International Nuclear Information System (INIS)

    2003-01-01

    The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases

  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. 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.

  6. 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

  7. 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

  8. Radial velocity observations of VB10

    Science.gov (United States)

    Deshpande, R.; Martin, E.; Zapatero Osorio, M. R.; Del Burgo, C.; Rodler, F.; Montgomery, M. M.

    2011-07-01

    VB 10 is the smallest star known to harbor a planet according to the recent astrometric study of Pravdo & Shaklan [1]. Here we present near-infrared (J-band) radial velocity of VB 10 performed from high resolution (R~20,000) spectroscopy (NIRSPEC/KECK II). Our results [2] suggest radial velocity variability with amplitude of ~1 km/s, a result that is consistent with the presence of a massive planet companion around VB10 as found via long-term astrometric monitoring of the star by Pravdo & Shaklan. Employing an entirely different technique we verify the results of Pravdo & Shaklan.

  9. Radial scar/complex sclerosing lesion of the breast--value of ultrasound.

    Science.gov (United States)

    Grunwald, S; Heyer, H; Kühl, A; Schwesinger, G; Schimming, A; Köhler, G; Ohlinger, R

    2007-04-01

    Although benign, radial scar/complex sclerosing adenosis is a lesion which histopathologically resembles tubular carcinoma. On physical examination, it is difficult to distinguish radial scar from a malignant tumour. Mammography cannot differentiate radial scar from malignancy. This clinical study aims to delineate the role of preoperative ultrasonography with emphasis on the question whether ultrasonography could lower the number of false-positive readings and therefore the number of open biopsies required. In this examination, we present the clinical, mammographic, ultrasonographic, and histopathological features of 6 cases of radial scars. Although most authors describe radial scars as non-palpable, 2 of 6 lesions were indeed palpable. On mammograms, radial scars have a spiculated appearance, a feature observed in all of our cases. Numerous ultrasonographic characteristics are listed in the literature, but ultrasonography is not reported to have clear-cut advantages. Although this study did not elucidate any unique ultrasonographic features to characterise these lesions, the analysis of all ultrasonographic results made us recognise a set of "nearly specific ultrasonographic features" of radial scars. Current B-mode imaging does not appear to lead to the desirable reduction of the rate of unnecessary open biopsies.

  10. 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

  11. 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.

  12. Development of a simplified calculational model for the transient core bowing effect

    International Nuclear Information System (INIS)

    Yokoo, Takeshi

    1997-01-01

    A simplified method to analyze the transient core radial deformation has been developed based on a model that calculates the shape of a single representative fuel assembly on the outermost row. The plant transient code CERES has been revised utilizing this method so that a integrated calculational process for the core neutronics, thermal-hydraulics and deformation can be realized. Using CERES, the responses of a 1000MWe class pool type metal fuel FBR plant during a ULOF event are calculated. According to the results, it is clarified that a passive shutdown without coolant boiling is attainable by selecting appropriate values for major design parameters such as the gap width between load-pad and the pad material properties. The maximum coolant temperature during ULOF is found to be 790C when the above core load-pad gap is set to 0.05 mm, which can be regarded as the most likely valued. The temperature increases to 915C but is still lower than the boiling point when 40% of uncertainty is taken into account. (author)

  13. 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.

  14. Large epidemic thresholds emerge in heterogeneous networks of heterogeneous nodes

    Science.gov (United States)

    Yang, Hui; Tang, Ming; Gross, Thilo

    2015-08-01

    One of the famous results of network science states that networks with heterogeneous connectivity are more susceptible to epidemic spreading than their more homogeneous counterparts. In particular, in networks of identical nodes it has been shown that network heterogeneity, i.e. a broad degree distribution, can lower the epidemic threshold at which epidemics can invade the system. Network heterogeneity can thus allow diseases with lower transmission probabilities to persist and spread. However, it has been pointed out that networks in which the properties of nodes are intrinsically heterogeneous can be very resilient to disease spreading. Heterogeneity in structure can enhance or diminish the resilience of networks with heterogeneous nodes, depending on the correlations between the topological and intrinsic properties. Here, we consider a plausible scenario where people have intrinsic differences in susceptibility and adapt their social network structure to the presence of the disease. We show that the resilience of networks with heterogeneous connectivity can surpass those of networks with homogeneous connectivity. For epidemiology, this implies that network heterogeneity should not be studied in isolation, it is instead the heterogeneity of infection risk that determines the likelihood of outbreaks.

  15. Large epidemic thresholds emerge in heterogeneous networks of heterogeneous nodes.

    Science.gov (United States)

    Yang, Hui; Tang, Ming; Gross, Thilo

    2015-08-21

    One of the famous results of network science states that networks with heterogeneous connectivity are more susceptible to epidemic spreading than their more homogeneous counterparts. In particular, in networks of identical nodes it has been shown that network heterogeneity, i.e. a broad degree distribution, can lower the epidemic threshold at which epidemics can invade the system. Network heterogeneity can thus allow diseases with lower transmission probabilities to persist and spread. However, it has been pointed out that networks in which the properties of nodes are intrinsically heterogeneous can be very resilient to disease spreading. Heterogeneity in structure can enhance or diminish the resilience of networks with heterogeneous nodes, depending on the correlations between the topological and intrinsic properties. Here, we consider a plausible scenario where people have intrinsic differences in susceptibility and adapt their social network structure to the presence of the disease. We show that the resilience of networks with heterogeneous connectivity can surpass those of networks with homogeneous connectivity. For epidemiology, this implies that network heterogeneity should not be studied in isolation, it is instead the heterogeneity of infection risk that determines the likelihood of outbreaks.

  16. Interpretation of active neutron measurements by the heterogeneous theory

    International Nuclear Information System (INIS)

    Birkhoff, G.; Depraz, J.; Descieux, J.P.

    1979-01-01

    In this paper are presented results from a study on the application of the heterogeneous method for the interpretation of active neutron measurements. The considered apparatus consists out of a cylindrical lead pile, which is provided with two axial channels: a central channel incorporates an antimony beryllium photoneutron source and an excentric channel serves for the insertion of the sample to be assayed for fissionable materials contents. The mathematical model of this apparatus is the heterogeneous group diffusion theory. Sample and source channel are described by multigroup monopolar and dipolar sources and sinks. Monopolar sources take account of neutron production within energy group and in-scatter from upper groups. Monopolar sinks represent neutron removal by absorption within energy group and outscatter to lower groups. Dipol sources describe radial streaming of neutrons across the sample channel. Multigroup diffusion theory is applied throughout the lead pile. The strengths of the monopolar and dipolar sources and sinks are determined by linear extrapolation distances of azimuthal mean and first harmonic flux values at the channels' surface. In an experiment we may measure the neutrons leaking out of the lead pile and linear extrapolation distances at the channels' surface. Such informations are utilized for interpretation in terms of fission neutron source strengh and mean neutron flux values in the sample. In this paper we summarized the theoretical work in course

  17. 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

  18. Fuel radial design using Path Relinking

    International Nuclear Information System (INIS)

    Campos S, Y.

    2007-01-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  19. Numerical Simulation of Hydraulic Fracture Propagation Guided by Single Radial Boreholes

    Directory of Open Access Journals (Sweden)

    Tiankui Guo

    2017-10-01

    Full Text Available Conventional hydraulic fracturing is not effective in target oil development zones with available wellbores located in the azimuth of the non-maximum horizontal in-situ stress. To some extent, we think that the radial hydraulic jet drilling has the function of guiding hydraulic fracture propagation direction and promoting deep penetration, but this notion currently lacks an effective theoretical support for fracture propagation. In order to verify the technology, a 3D extended finite element numerical model of hydraulic fracturing promoted by the single radial borehole was established, and the influences of nine factors on propagation of hydraulic fracture guided by the single radial borehole were comprehensively analyzed. Moreover, the term ‘Guidance factor (Gf’ was introduced for the first time to effectively quantify the radial borehole guidance. The guidance of nine factors was evaluated through gray correlation analysis. The experimental results were consistent with the numerical simulation results to a certain extent. The study provides theoretical evidence for the artificial control technology of directional propagation of hydraulic fracture promoted by the single radial borehole, and it predicts the guidance effect of a single radial borehole on hydraulic fracture to a certain extent, which is helpful for planning well-completion and fracturing operation parameters in radial borehole-promoted hydraulic fracturing technology.

  20. Radial collective flow in heavy-ion collisions at intermediate energies

    International Nuclear Information System (INIS)

    Borderie, B.

    1996-11-01

    The production of radial collective flow is associated with collisions leading to sources which undergo multifragmentation/explosion processes. After a theoretical survey of possible causes of production of radial flow, methods used to derive experimental values are discussed. Finally, a large set of data is presented which can be used to study and disentangle the different effects leading to radial collective flow. The dominant role of compression in the lower energy domain is emphasized. (author)

  1. Rotary and radial forcing effects on center-of-mass locomotion dynamics.

    Science.gov (United States)

    Shen, Z H; Larson, P L; Seipel, J E

    2014-09-01

    Rotary and radial forcing are two common actuation methods for legged robots. However, these two orthogonal methods of center-of-mass (CoM) forcing have not been compared as potentially alternative strategies of actuation. In this paper, we compare the CoM stability and energetics of running with rotary and radial actuation through the simulation of two models: the rotary-forced spring-loaded inverted pendulum (rotary-forced-SLIP), and the radially-forced-SLIP. We model both radial and rotary actuation in the simplest way, applying them as a constant force during the stance portion of the gait. A simple application of constant rotary forcing throughout stance is capable of producing fully-asymptotically stable motion; however, a similarly constant application of radial forcing throughout the stance is not capable of producing stable solutions. We then allow both the applied rotary and radial forcing functions to turn on or off based on the occurrence of the mid-stance event, which breaks the symmetry of actuation during stance towards a net forward propulsion. We find that both a rotary force applied in the first half of stance and a radial force applied in the second half of stance, are capable of stabilizing running. Interestingly, these two forcing methods improve the motion stability in different ways. Rotary forcing first reduces then greatly increases the size of the stable parameter region when gradually increased. Radial forcing expands the stable parameter region, but only in a moderate way. Also, it is found that parameter region stabilized by rotary and radial forcing are largely complementary. Overall, rotary forcing can better stabilize running for both constant and event-based forcing functions that were attempted. This indicates that rotary forcing has an inherent capability of stabilizing running, even when minimal time-or-event-or-state feedback is present. Radial forcing, however, tends to be more energy efficient when compared to rotary forcing

  2. Fast radial basis functions for engineering applications

    CERN Document Server

    Biancolini, Marco Evangelos

    2017-01-01

    This book presents the first “How To” guide to the use of radial basis functions (RBF). It provides a clear vision of their potential, an overview of ready-for-use computational tools and precise guidelines to implement new engineering applications of RBF. Radial basis functions (RBF) are a mathematical tool mature enough for useful engineering applications. Their mathematical foundation is well established and the tool has proven to be effective in many fields, as the mathematical framework can be adapted in several ways. A candidate application can be faced considering the features of RBF:  multidimensional space (including 2D and 3D), numerous radial functions available, global and compact support, interpolation/regression. This great flexibility makes RBF attractive – and their great potential has only been partially discovered. This is because of the difficulty in taking a first step toward RBF as they are not commonly part of engineers’ cultural background, but also due to the numerical complex...

  3. On improved confinement in mirror plasmas by a radial electric field

    Science.gov (United States)

    Ågren, O.; Moiseenko, V. E.

    2017-11-01

    A weak radial electric field can suppress radial excursions of a guiding center from its mean magnetic surface. The physical origin of this effect is the smearing action by a poloidal E × B rotation, which tend to cancel out the inward and outward radial drifts. A use of this phenomenon may provide larger margins for magnetic field shaping with radial confinement of particles maintained in the collision free idealization. Mirror fields, stabilized by a quadrupolar field component, are of particular interest for their MHD stability and the possibility to control the quasi neutral radial electric field by biased potential plates outside the confinement region. Flux surface footprints on the end tank wall have to be traced to avoid short-circuiting between biased plates. Assuming a robust biasing procedure, moderate voltage demands for the biased plates seems adequate to cure even the radial excursions of Yushmanov ions which could be locally trapped near the mirrors. Analytical expressions are obtained for a magnetic quadrupolar mirror configuration which possesses minimal radial magnetic drifts in the central confinement region. By adding a weak controlled radial quasi-neutral electric field, the majority of gyro centers are predicted to be forced to move even closer to their respective mean magnetic surface. The gyro center radial coordinate is in such a case an accurate approximation for a constant of motion. By using this constant of motion, the analysis is in a Vlasov description extended to finite β. A correspondence between that Vlasov system and a fluid description with a scalar pressure and an electric potential is verified. The minimum B criterion is considered and implications for flute mode stability in the considered magnetic field is analyzed. By carrying out a long-thin expansion to a higher order, the validity of the calculations are extended to shorter and more compact device designs.

  4. The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

    International Nuclear Information System (INIS)

    Williams, M. G.; Mouser, M. R.; Simon, J. B.

    2012-01-01

    The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

  5. 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.)

  6. Radial Transport and Meridional Circulation in Accretion Disks

    Energy Technology Data Exchange (ETDEWEB)

    Philippov, Alexander A. [Department of Astrophysical Sciences, Princeton University, Ivy Lane, Princeton, NJ 08540 (United States); Rafikov, Roman R., E-mail: sashaph@princeton.edu [Institute for Advanced Study, Einstein Drive, Princeton, NJ 08540 (United States)

    2017-03-10

    Radial transport of particles, elements and fluid driven by internal stresses in three-dimensional (3D) astrophysical accretion disks is an important phenomenon, potentially relevant for the outward dust transport in protoplanetary disks, origin of the refractory particles in comets, isotopic equilibration in the Earth–Moon system, etc. To gain better insight into these processes, we explore the dependence of meridional circulation in 3D disks with shear viscosity on their thermal stratification, and demonstrate a strong effect of the latter on the radial flow. Previous locally isothermal studies have normally found a pattern of the radial outflow near the midplane, switching to inflow higher up. Here we show, both analytically and numerically, that a flow that is inward at all altitudes is possible in disks with entropy and temperature steeply increasing with height. Such thermodynamic conditions may be typical in the optically thin, viscously heated accretion disks. Disks in which these conditions do not hold should feature radial outflow near the midplane, as long as their internal stress is provided by the shear viscosity. Our results can also be used for designing hydrodynamical disk simulations with a prescribed pattern of the meridional circulation.

  7. Detonation in supersonic radial outflow

    KAUST Repository

    Kasimov, Aslan R.; Korneev, Svyatoslav

    2014-01-01

    We report on the structure and dynamics of gaseous detonation stabilized in a supersonic flow emanating radially from a central source. The steady-state solutions are computed and their range of existence is investigated. Two-dimensional simulations

  8. 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.)

  9. ACR-1000: Operator - based development

    International Nuclear Information System (INIS)

    Shalaby, B.; Alizadeh, A.

    2007-01-01

    Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU * reactors to establish Generation III + Advanced CANDU Reactor T M (ACR T M) technology. The ACR-1000 T M nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDU T M technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of

  10. Mirror Advanced Reactor Study (MARS)

    International Nuclear Information System (INIS)

    Logan, B.G.

    1983-01-01

    Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made

  11. THE OCCURRENCE OF THE RADIAL CLUB HAND IN CHILDREN WITH DIFFERENT SYNDROMES

    Directory of Open Access Journals (Sweden)

    Sergey Ivanovich Golyana

    2013-03-01

    Full Text Available Radial club hand is a developmental anomaly of the upper extremity, being characterized as a longitudinal underdevelopment of a forearm and a hand on the radial surface, consisting in a hypo-/ aplazy radial bone and the thumb of various degree of expressiveness. Characteristic symptoms of this developmental anomaly are: shortening and bow-shaped curvature of a forearm, palmar and radial deviation of a hand, underdevelopment of the thumb from its proximal departments and structures, anomaly of development of three-phalanx fingers of a hand (is more often than the 2-4th, violation of a cosmetic condition and functionality of the affected segment. From 2000 for 2012 in FSI SRICO n.a. H.Turner examination and treatment of 23 children with various syndromes at which the radial club hand was revealed are conducted. The main syndromes at which it is revealed radial club hand - Holt-Orama syndrome, TAR- syndrome and VACTERL syndrome. Tactics and techniques of surgical treatment of a radial club hand it various syndromes most often don’t differ from treatment of other types of a radial club hand though demand an individual approach depending on severity and a type of deformation of the upper extremity.

  12. The effect of radial migration on galactic disks

    International Nuclear Information System (INIS)

    Vera-Ciro, Carlos; D'Onghia, Elena; Navarro, Julio; Abadi, Mario

    2014-01-01

    We study the radial migration of stars driven by recurring multi-arm spiral features in an exponential disk embedded in a dark matter halo. The spiral perturbations redistribute angular momentum within the disk and lead to substantial radial displacements of individual stars, in a manner that largely preserves the circularity of their orbits and that results, after 5 Gyr (∼40 full rotations at the disk scale length), in little radial heating and no appreciable changes to the vertical or radial structure of the disk. Our results clarify a number of issues related to the spatial distribution and kinematics of migrators. In particular, we find that migrators are a heavily biased subset of stars with preferentially low vertical velocity dispersions. This 'provenance bias' for migrators is not surprising in hindsight, for stars with small vertical excursions spend more time near the disk plane, and thus respond more readily to non-axisymmetric perturbations. We also find that the vertical velocity dispersion of outward migrators always decreases, whereas the opposite holds for inward migrators. To first order, newly arrived migrators simply replace stars that have migrated off to other radii, thus inheriting the vertical bias of the latter. Extreme migrators might therefore be recognized, if present, by the unexpectedly small amplitude of their vertical excursions. Our results show that migration, understood as changes in angular momentum that preserve circularity, can strongly affect the thin disk, but cast doubts on models that envision the Galactic thick disk as a relic of radial migration.

  13. Rayleigh-Taylor instability of cylindrical jets with radial motion

    International Nuclear Information System (INIS)

    Chen, X.M.; Schrock, V.E.; Peterson, P.F.

    1997-01-01

    Rayleigh-Taylor instability of an interface between fluids with different densities subjected to acceleration normal to itself has interested researchers for almost a century. The classic analyses of a flat interface by Rayleigh and Taylor have shown that this type of instability depends on the direction of acceleration and the density differences of the two fluids. Plesset later analyzed the stability of a spherically symmetric flows (and a spherical interface) and concluded that the instability also depends on the velocity of the interface as well as the direction and magnitude of radial acceleration. The instability induced by radial motion in cylindrical systems seems to have been neglected by previous researchers. This paper analyzes the Rayleigh-Taylor type of instability for a cylindrical surface with radial motions. The results of the analysis show that, like the spherical case, the radial velocity also plays an important role. As an application, the example of a liquid jet surface in an Inertial Confinement Fusion (ICF) reactor design is analyzed. (orig.)

  14. A Novel Integrated Structure with a Radial Displacement Sensor and a Permanent Magnet Biased Radial Magnetic Bearing

    Directory of Open Access Journals (Sweden)

    Jinji Sun

    2014-01-01

    Full Text Available In this paper, a novel integrated structure is proposed in order to reduce the axial length of the high speed of a magnetically suspended motor (HSMSM to ensure the maximum speed, which combines radial displacement sensor probes and the permanent magnet biased radial magnetic bearing in HSMSM. The sensor probes are integrated in the magnetic bearing, and the sensor preamplifiers are placed in the control system of the HSMSM, separate from the sensor probes. The proposed integrated structure can save space in HSMSMs, improve the working frequency, reduce the influence of temperature on the sensor circuit, and improve the stability of HSMSMs.

  15. Radial-piston pump for drive of test machines

    Science.gov (United States)

    Nizhegorodov, A. I.; Gavrilin, A. N.; Moyzes, B. B.; Cherkasov, A. I.; Zharkevich, O. M.; Zhetessova, G. S.; Savelyeva, N. A.

    2018-01-01

    The article reviews the development of radial-piston pump with phase control and alternating-flow mode for seismic-testing platforms and other test machines. The prospects for use of the developed device are proved. It is noted that the method of frequency modulation with the detection of the natural frequencies is easily realized by using the radial-piston pump. The prospects of further research are given proof.

  16. 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)

  17. Nuclear Option in Korea

    International Nuclear Information System (INIS)

    Han, K. I.

    2002-01-01

    With sixteen(16) operating nuclear units in Korea, the share of nuclear power generation reached 41% of the total electric power generation as of December 2000. A prediction is that it would further increase to 44.5% by year 2015 according to the national long term power development plan. Four units are currently under construction with 6 more units in order. With little domestic energy resource and increasing energy demand to support national economic growth, Korea has chosen nuclear power as one of the major energy sources to ensure stable power supply and to promote energy self-sufficiency. It has been recognized that nuclear power in Korea is not a selective option but rather a necessity. The Korean nuclear power development started with construction of a 600 MWe size reactor that was designed and constructed by foreign vendors. As the national grid capacity became larger, the size of nuclear units increased to 1000 MWe class. In the mean time, the need for nuclear technology self-reliance grew not only in operation and maintenance but also in construction, manufacturing and design. For this, a nuclear technology self-reliance program has been embarked with the support of the Government and utility, and the 1000 MWe class KSNP(Korean Standard Nuclear Power Plant) has been developed. The KSNPs are currently being designed, manufactured, constructed and operated by relevant Korean entities themselves. To fit into a larger capacity national grid and also to improve nuclear economic competitiveness, the 1400 MWe class KNGR(Korean Next Generation Reactor) design has been developed uprating the 1000 MWe KSNP design. Its construction project is currently under contract negotiation, and is planned to be finished by 2010. In the mean time, to be ready for future electric power market deregulation, the 600 MWe class small KSNP design is being developed downsizing the KSNP. A modular small size reactor, SMART(System Integrated Modular Advanced Reactor) is also being

  18. SpicyNodes Radial Map Engine

    Science.gov (United States)

    Douma, M.; Ligierko, G.; Angelov, I.

    2008-10-01

    The need for information has increased exponentially over the past decades. The current systems for constructing, exploring, classifying, organizing, and searching information face the growing challenge of enabling their users to operate efficiently and intuitively in knowledge-heavy environments. This paper presents SpicyNodes, an advanced user interface for difficult interaction contexts. It is based on an underlying structure known as a radial map, which allows users to manipulate and interact in a natural manner with entities called nodes. This technology overcomes certain limitations of existing solutions and solves the problem of browsing complex sets of linked information. SpicyNodes is also an organic system that projects users into a living space, stimulating exploratory behavior and fostering creative thought. Our interactive radial layout is used for educational purposes and has the potential for numerous other applications.

  19. Modelling and analysis of radial thermal stresses and temperature ...

    African Journals Online (AJOL)

    A theoretical investigation has been undertaken to study operating temperatures, heat fluxes and radial thermal stresses in the valves of a modern diesel engine with and without air-cavity. Temperatures, heat fluxes and radial thermal stresses were measured theoretically for both cases under all four thermal loading ...

  20. 7 CFR 1000.86 - Deduction for marketing services.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 9 2010-01-01 2009-01-01 true Deduction for marketing services. 1000.86 Section 1000... SERVICE (Marketing Agreements and Orders; Milk), DEPARTMENT OF AGRICULTURE GENERAL PROVISIONS OF FEDERAL MILK MARKETING ORDERS Administrative Assessment and Marketing Service Deduction § 1000.86 Deduction for...

  1. Constraints on small-scale heterogeneity in the lowermost mantle from observations of near podal PcP precursors

    Science.gov (United States)

    Zhang, Baolong; Ni, Sidao; Sun, Daoyuan; Shen, Zhichao; Jackson, Jennifer M.; Wu, Wenbo

    2018-05-01

    Volumetric heterogeneities on large (∼>1000 km) and intermediate scales (∼>100 km) in the lowermost mantle have been established with seismological approaches. However, there are controversies regarding the level of heterogeneity in the lowermost mantle at small scales (a few kilometers to tens of kilometers), with lower bound estimates ranging from 0.1% to a few percent. We take advantage of the small amplitude PcP waves at near podal distances (0-12°) to constrain the level of small-scale heterogeneity within 250 km above the CMB. First, we compute short period synthetic seismograms with a finite difference code for a series of volumetric heterogeneity models in the lowermost mantle, and find that PcP is not identifiable if the small-scale heterogeneity in the lowermost mantle is above 2.5%. We then use a functional form appropriate for coda decay to suppress P coda contamination. By comparing the corrected envelope of PcP and its precursors with synthetic seismograms, we find that perturbations of small-scale (∼8 km) heterogeneity in the lowermost mantle is ∼0.2-0.5% beneath regions of the China-Myanmar border area, Okhotsk Sea and South America. Whereas strong perturbations (∼1.0%) are found beneath Central America. In the regions studied, we find that this particular type of small-scale heterogeneity in the lowermost mantle is weak, yet there are some regions requiring heterogeneity up to 1.0%. Where scattering is stronger, such as under Central America, more chemically complex mineral assemblages may be present at the core-mantle boundary.

  2. A novel integrated 4-DOF radial hybrid magnetic bearing for MSCMG

    Energy Technology Data Exchange (ETDEWEB)

    Jinji, Sun; Ziyan, Ju [School of Instrumentation Science & Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, Science and Technology on Inertial Laboratory, Beijing 100191 (China); Weitao, Han, E-mail: hanweitaotao@163.com [CRRC Qingdao Sifang CO., LTD, Qingdao 266111 (China); Gang, Liu [School of Instrumentation Science & Opto-electronics Engineering, Beijing University of Aeronautics and Astronautics, Science and Technology on Inertial Laboratory, Beijing 100191 (China)

    2017-01-01

    This paper proposes a novel integrated radial hybrid magnetic bearing (RHMB) for application with the small-sized magnetically suspended control moment gyroscope (MSCMG), which can control four degrees of freedom (4-DOFs), including two radial translational DOFs and two radial tilting DOFs, and provide the axial passive resilience. The configuration and working principle of the RHMB are introduced. Mathematical models of radial force, axial resilience and moment are established by using equivalent magnetic circuit method (EMCM), from which the radial force–radial displacement, radial force–current relationships are derived, as well as axial resilience–axial displacement, moment–tilting angle and moment–current. Finite element method (FEM) is also applied to analyze the performance and characteristics of the RHMB. The analysis results are in good agreement with that calculated by the EMCM, which is helpful in designing, optimizing and controlling the RHMB. The comparisons between the performances of the integrated 4-DOF RHMB and the traditional 4-DOF RHMB are made. The contrast results indicate that the proposed integrated 4-DOF RHMB possesses better performance compared to the traditional structure, such as copper loss, current stiffness, and tilting current stiffness. - Highlights: • An integrated 4-DOF RHMB is proposed for the small-sized MSCMG. • The 4-DOF RHMB has good linear force–displacement and force–current characteristics. • The RHMB has good linear moment–current and the moment–tilting angle characteristic.

  3. 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.)

  4. 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

  5. MRI of radial displacement of the meniscus in the knee

    International Nuclear Information System (INIS)

    Chen Jian; Lv Houshan; Lao Shan; Guan Zhenpeng; Hong Nan; Liang Hao

    2006-01-01

    Objective: To describe the phenomenon of radial displacement of the meniscus of the knees in the study population with MR imaging, and to establish MRI diagnostic criteria for radial displacement of the meniscus and displacement index. Methods: MR signs of radial displacement of the meniscus were evaluated retrospectively in 398 patients with knee symptoms who were examined with non- weight bearing MR images from Jan. 2000 to Feb. 2004. The patients younger than 18 years old, with joint effusion or serious arthropathy were excluded and 312 patients were eligible to be enrolled in this study. The criterion for radial displacement of the meniscus was defined as the location of the edge of meniscal body beyond the femoral and tibial outer border line. A displacement index, defined as the ratio of meniscal overhang to meniscal width, was used to quantify meniscal displacement. Results: The prevalence of radial displacement of the meniscus was 16.7% (52/312) and 13.9% (21/151) in right knee and 19.3% (31/161 )in left knee, respectively. There was no significant difference between left and right knee (χ 2 =1.60, P>0.05) and the ratio between medial and lateral meniscus was 7.8:1. The average displacement index was 0.54±0.24. The displacement indices were significant higher in older group (F=3.63, P<0.05). The incidence and indices of radial displacement of the meniscus for patients under or above 50 year older were 12.0%(17/142), 0.46±0.22 and 20.6% (35/170), 0.64±0.20, respectively. Difference was highly significant (t=0.84, P<0.01). Conclusion: It was concluded that radial displacement of the meniscus in knees was not a rare finding with MR imaging in patients with knee symptoms. The incidence increased in older age group. Further investigations were recommended to understand the etiology and clinical significance of the phenomenon of radial displacement of the meniscus. (authors)

  6. 21 CFR 892.1000 - Magnetic resonance diagnostic device.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Magnetic resonance diagnostic device. 892.1000 Section 892.1000 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES RADIOLOGY DEVICES Diagnostic Devices § 892.1000 Magnetic resonance diagnostic...

  7. 42 CFR 1000.20 - Definitions specific to Medicare.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 5 2010-10-01 2010-10-01 false Definitions specific to Medicare. 1000.20 Section 1000.20 Public Health OFFICE OF INSPECTOR GENERAL-HEALTH CARE, DEPARTMENT OF HEALTH AND HUMAN SERVICES GENERAL PROVISIONS INTRODUCTION; GENERAL DEFINITIONS Definitions § 1000.20 Definitions specific to...

  8. 13 CFR 120.1000 - Risk-Based Lender Oversight.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Risk-Based Lender Oversight. 120.1000 Section 120.1000 Business Credit and Assistance SMALL BUSINESS ADMINISTRATION BUSINESS LOANS Risk-Based Lender Oversight Supervision § 120.1000 Risk-Based Lender Oversight. (a) Risk-Based Lender...

  9. Vortex Whistle in Radial Intake

    National Research Council Canada - National Science Library

    Tse, Man-Chun

    2004-01-01

    In a radial-to-axial intake with inlet guide vanes (IGV) at the entry, a strong flow circulation Gamma can be generated from the tangential flow components created by the IGVs when their setting exceed about halfclosing (approx. 45 deg...

  10. Radial force measurement of endovascular stents: Influence of stent design and diameter.

    Science.gov (United States)

    Matsumoto, Takuya; Matsubara, Yutaka; Aoyagi, Yukihiko; Matsuda, Daisuke; Okadome, Jun; Morisaki, Koichi; Inoue, Kentarou; Tanaka, Shinichi; Ohkusa, Tomoko; Maehara, Yoshihiko

    2016-04-01

    Angioplasty and endovascular stent placement is used in case to rescue the coverage of main branches to supply blood to brain from aortic arch in thoracic endovascular aortic repair. This study assessed mechanical properties, especially differences in radial force, of different endovascular and thoracic stents. We analyzed the radial force of three stent models (Epic, E-Luminexx and SMART) stents using radial force-tester method in single or overlapping conditions. We also analyzed radial force in three thoracic stents using Mylar film testing method: conformable Gore-TAG, Relay, and Valiant Thoracic Stent Graft. Overlapping SMART stents had greater radial force than overlapping Epic or Luminexx stents (P stents was greater than that of all three endovascular stents (P stents, site of deployment, and layer characteristics. In clinical settings, an understanding of the mechanical characteristics, including radial force, is important in choosing a stent for each patient. © The Author(s) 2015.

  11. Organized proteomic heterogeneity in colorectal cancer liver metastases and implications for therapies.

    Science.gov (United States)

    Turtoi, Andrei; Blomme, Arnaud; Debois, Delphine; Somja, Joan; Delvaux, David; Patsos, Georgios; Di Valentin, Emmanuel; Peulen, Olivier; Mutijima, Eugène Nzaramba; De Pauw, Edwin; Delvenne, Philippe; Detry, Olivier; Castronovo, Vincent

    2014-03-01

    Tumor heterogeneity is a major obstacle for developing effective anticancer treatments. Recent studies have pointed to large stochastic genetic heterogeneity within cancer lesions, where no pattern seems to exist that would enable a more structured targeted therapy approach. Because to date no similar information is available at the protein (phenotype) level, we employed matrix assisted laser desorption ionization (MALDI) image-guided proteomics and explored the heterogeneity of extracellular and membrane subproteome in a unique collection of eight fresh human colorectal carcinoma (CRC) liver metastases. Monitoring the spatial distribution of over 1,000 proteins, we found unexpectedly that all liver metastasis lesions displayed a reproducible, zonally delineated pattern of functional and therapeutic biomarker heterogeneity. The peritumoral region featured elevated lipid metabolism and protein synthesis, the rim of the metastasis displayed increased cellular growth, movement, and drug metabolism, whereas the center of the lesion was characterized by elevated carbohydrate metabolism and DNA-repair activity. From the aspect of therapeutic targeting, zonal expression of known and novel biomarkers was evident, reinforcing the need to select several targets in order to achieve optimal coverage of the lesion. Finally, we highlight two novel antigens, LTBP2 and TGFBI, whose expression is a consistent feature of CRC liver metastasis. We demonstrate their in vivo antibody-based targeting and highlight their potential usefulness for clinical applications. The proteome heterogeneity of human CRC liver metastases has a distinct, organized pattern. This particular hallmark can now be used as part of the strategy for developing rational therapies based on multiple sets of targetable antigens. © 2014 by the American Association for the Study of Liver Diseases.

  12. 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.

  13. PROLONGED RADIAL ARTERY SPASM IN THE CATHETERIZATION LABORATORY - RELIEF BY PHARMACOLOGICAL INTERVENTION

    Directory of Open Access Journals (Sweden)

    Krishna Kumar

    2010-11-01

    Full Text Available Radial spasm is often very prolonged and painful to the patient. Here, we describe a novel way to deal with the same. The total spasm lasted over 4 hours. A 3.4 6 JR catheter was introduced via the femoral route and papav arine one ampoule was injected directly into the right subclavian artery. After about 10 min we were able to pull out the radial catheter. Radial angiography is a simple procedure with reportedly less complications 1,2. How ever ,it has one major complication radial spasm. We describe here a patient with radial spasm that persisted for more than 2 hours and how we dealt with it.

  14. Rotary and radial forcing effects on center-of-mass locomotion dynamics

    International Nuclear Information System (INIS)

    Shen, Z H; Larson, P L; Seipel, J E

    2014-01-01

    Rotary and radial forcing are two common actuation methods for legged robots. However, these two orthogonal methods of center-of-mass (CoM) forcing have not been compared as potentially alternative strategies of actuation. In this paper, we compare the CoM stability and energetics of running with rotary and radial actuation through the simulation of two models: the rotary-forced spring-loaded inverted pendulum (rotary-forced-SLIP), and the radially-forced-SLIP. We model both radial and rotary actuation in the simplest way, applying them as a constant force during the stance portion of the gait. A simple application of constant rotary forcing throughout stance is capable of producing fully-asymptotically stable motion; however, a similarly constant application of radial forcing throughout the stance is not capable of producing stable solutions. We then allow both the applied rotary and radial forcing functions to turn on or off based on the occurrence of the mid-stance event, which breaks the symmetry of actuation during stance towards a net forward propulsion. We find that both a rotary force applied in the first half of stance and a radial force applied in the second half of stance, are capable of stabilizing running. Interestingly, these two forcing methods improve the motion stability in different ways. Rotary forcing first reduces then greatly increases the size of the stable parameter region when gradually increased. Radial forcing expands the stable parameter region, but only in a moderate way. Also, it is found that parameter region stabilized by rotary and radial forcing are largely complementary. Overall, rotary forcing can better stabilize running for both constant and event-based forcing functions that were attempted. This indicates that rotary forcing has an inherent capability of stabilizing running, even when minimal time-or-event-or-state feedback is present. Radial forcing, however, tends to be more energy efficient when compared to rotary forcing

  15. A Novel Repair Method for Radial Tears of the Medial Meniscus: Biomechanical Comparison of Transtibial 2-Tunnel and Double Horizontal Mattress Suture Techniques Under Cyclic Loading.

    Science.gov (United States)

    Bhatia, Sanjeev; Civitarese, David M; Turnbull, Travis Lee; LaPrade, Christopher M; Nitri, Marco; Wijdicks, Coen A; LaPrade, Robert F

    2016-03-01

    Complete radial tears of the medial meniscus have been reported to be functionally similar to a total meniscectomy. At present, there is no consensus on an ideal technique for repair of radial midbody tears of the medial meniscus. Prior attempts at repair with double horizontal mattress suture techniques have led to a reportedly high rate of incomplete healing or healing in a nonanatomic (gapped) position, which compromises the ability of the meniscus to withstand hoop stresses. A newly proposed 2-tunnel radial meniscal repair method will result in decreased gapping and increased ultimate failure loads compared with the double horizontal mattress suture repair technique under cyclic loading. Controlled laboratory study. Ten matched pairs of male human cadaveric knees (average age, 58.6 years; range, 48-66 years) were used. A complete radial medial meniscal tear was made at the junction of the posterior one-third and middle third of the meniscus. One knee underwent a horizontal mattress inside-out repair, while the contralateral knee underwent a radial meniscal repair entailing the same technique with a concurrent novel 2-tunnel repair. Specimens were potted and mounted on a universal testing machine. Each specimen was cyclically loaded 1000 times with loads between 5 and 20 N before experiencing a load to failure. Gap distances at the tear site and failure load were measured. The 2-tunnel repairs exhibited a significantly stronger ultimate failure load (median, 196 N; range, 163-212 N) than did the double horizontal mattress suture repairs (median, 106 N; range, 63-229 N) (P = .004). In addition, the 2-tunnel repairs demonstrated decreased gapping at all testing states (P meniscus significantly decrease the ability of the meniscus to dissipate tibiofemoral loads, predisposing patients to early osteoarthritis. Improving the ability to repair medial meniscal radial tears in a way that withstands cyclic loads and heals in an anatomic position could significantly

  16. The First Experience of Triple Nerve Transfer in Proximal Radial Nerve Palsy.

    Science.gov (United States)

    Emamhadi, Mohammadreza; Andalib, Sasan

    2018-01-01

    Injury to distal portion of posterior cord of brachial plexus leads to palsy of radial and axillary nerves. Symptoms are usually motor deficits of the deltoid muscle; triceps brachii muscle; and extensor muscles of the wrist, thumb, and fingers. Tendon transfers, nerve grafts, and nerve transfers are options for surgical treatment of proximal radial nerve palsy to restore some motor functions. Tendon transfer is painful, requires a long immobilization, and decreases donor muscle strength; nevertheless, nerve transfer produces promising outcomes. We present a patient with proximal radial nerve palsy following a blunt injury undergoing triple nerve transfer. The patient was involved in a motorcycle accident with complete palsy of the radial and axillary nerves. After 6 months, on admission, he showed spontaneous recovery of axillary nerve palsy, but radial nerve palsy remained. We performed triple nerve transfer, fascicle of ulnar nerve to long head of the triceps branch of radial nerve, flexor digitorum superficialis branch of median nerve to extensor carpi radialis brevis branch of radial nerve, and flexor carpi radialis branch of median nerve to posterior interosseous nerve, for restoration of elbow, wrist, and finger extensions, respectively. Our experience confirmed functional elbow, wrist, and finger extensions in the patient. Triple nerve transfer restores functions of the upper limb in patients with debilitating radial nerve palsy after blunt injuries. Copyright © 2017 Elsevier Inc. All rights reserved.

  17. Scandinavia 1000-1750

    DEFF Research Database (Denmark)

    Rasmussen, Carsten Porskrog; Myking, John Ragnar

    2010-01-01

    The article describes the distribution of property and power in rural Scandinavia (Denmark, Sweden and Norway) a.D. 1000-1750. It covers the distribution of property and structure of manors and holdings. It also looks at the structure of the peasantry, village communities, and the use of communal...

  18. Quantum Heterogeneous Computing for Satellite Positioning Optimization

    Science.gov (United States)

    Bass, G.; Kumar, V.; Dulny, J., III

    2016-12-01

    Hard optimization problems occur in many fields of academic study and practical situations. We present results in which quantum heterogeneous computing is used to solve a real-world optimization problem: satellite positioning. Optimization problems like this can scale very rapidly with problem size, and become unsolvable with traditional brute-force methods. Typically, such problems have been approximately solved with heuristic approaches; however, these methods can take a long time to calculate and are not guaranteed to find optimal solutions. Quantum computing offers the possibility of producing significant speed-up and improved solution quality. There are now commercially available quantum annealing (QA) devices that are designed to solve difficult optimization problems. These devices have 1000+ quantum bits, but they have significant hardware size and connectivity limitations. We present a novel heterogeneous computing stack that combines QA and classical machine learning and allows the use of QA on problems larger than the quantum hardware could solve in isolation. We begin by analyzing the satellite positioning problem with a heuristic solver, the genetic algorithm. The classical computer's comparatively large available memory can explore the full problem space and converge to a solution relatively close to the true optimum. The QA device can then evolve directly to the optimal solution within this more limited space. Preliminary experiments, using the Quantum Monte Carlo (QMC) algorithm to simulate QA hardware, have produced promising results. Working with problem instances with known global minima, we find a solution within 8% in a matter of seconds, and within 5% in a few minutes. Future studies include replacing QMC with commercially available quantum hardware and exploring more problem sets and model parameters. Our results have important implications for how heterogeneous quantum computing can be used to solve difficult optimization problems in any

  19. Development of a Radial Deconsolidation Method

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Fred C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    A series of experiments have been initiated to determine the retention or mobility of fission products* in AGR fuel compacts [Petti, et al. 2010]. This information is needed to refine fission product transport models. The AGR-3/4 irradiation test involved half-inch-long compacts that each contained twenty designed-to-fail (DTF) particles, with 20-μm thick carbon-coated kernels whose coatings were deliberately fabricated such that they would crack under irradiation, providing a known source of post-irradiation isotopes. The DTF particles in these compacts were axially distributed along the compact centerline so that the diffusion of fission products released from the DTF kernels would be radially symmetric [Hunn, et al. 2012; Hunn et al. 2011; Kercher, et al. 2011; Hunn, et al. 2007]. Compacts containing DTF particles were irradiated at Idaho National Laboratory (INL) at the Advanced Test Reactor (ATR) [Collin, 2015]. Analysis of the diffusion of these various post-irradiation isotopes through the compact requires a method to radially deconsolidate the compacts so that nested-annular volumes may be analyzed for post-irradiation isotope inventory in the compact matrix, TRISO outer pyrolytic carbon (OPyC), and DTF kernels. An effective radial deconsolidation method and apparatus appropriate to this application has been developed and parametrically characterized.

  20. A visual study of radial inward choked flow of liquid nitrogen.

    Science.gov (United States)

    Hendricks, R. C.; Simoneau, R. J.; Hsu, Y. Y.

    1973-01-01

    Data and high speed movies were acquired on pressurized subcooled liquid nitrogen flowing radially inward through a 0.0076 cm gap. The stagnation pressure ranged from 0.7 to 4 MN/sq m. Steady radial inward choked flow appears equivalent to steady choked flow through axisymmetric nozzles. Transient choked flows through the radial gap are not uniform and the discharge pattern appears as nonuniform impinging jets. The critical mass flow rate data for the transient case appear different from those for the steady case. On the mass flow rate vs pressure map, the slope and separation of the isotherms appear to be less for transient than for steady radial choked flow.

  1. Design and neutronic investigation of the Nano fluids application to VVER-1000 nuclear reactor with dual cooled annular fuel

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Ebrahimian, M.

    2016-01-01

    Highlights: • The change in neutronic parameters to the use of nanofluid as coolant is presented. • Nanoparticle deposition on fuel clad is investigated. • Radial and axial local power peaking factors are presented. • ZrO 2 and Al 2 O 3 have the lowest rate of K eff drop off. - Abstract: Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and K eff depression occurred and ZrO 2 and Al 2 O 3 have the lowest rate of drop off.

  2. Experimental study of the large-scale axially heterogeneous liquid-metal fast breeder reactor at the fast critical assembly: Power distribution measurements and their analyses

    International Nuclear Information System (INIS)

    Iijima, S.; Obu, M.; Hayase, T.; Ohno, A.; Nemoto, T.; Okajima, S.

    1988-01-01

    Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum harder in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket

  3. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  4. 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.

  5. Mejoramiento de imágenes usando funciones de base radial Images improvement using radial basis functions

    Directory of Open Access Journals (Sweden)

    Jaime Alberto Echeverri Arias

    2009-07-01

    Full Text Available La eliminación del ruido impulsivo es un problema clásico del procesado no lineal para el mejoramiento de imágenes y las funciones de base radial de soporte global son útiles para enfrentarlo. Este trabajo presenta una técnica de interpolación que disminuye eficientemente el ruido impulsivo en imágenes, mediante el uso de interpolante obtenido por funciones de base radial en el marco de la investigación enfocada en el desarrollo de un Sistema de recuperación de imágenes de recursos acuáticos amazónicos. Esta técnica primero etiqueta los píxeles de la imagen que son ruidosos y, mediante la interpolación, genera un valor de reconstrucción de dicho píxel usando sus vecinos. Los resultados obtenidos son comparables y muchas veces mejores que otras técnicas ya publicadas y reconocidas. Según el análisis de resultados, se puede aplicar a imágenes con altas tasas de ruido, manteniendo un bajo error de reconstrucción de los píxeles "ruidosos", así como la calidad visual.Global support radial base functions are effective in eliminating impulsive noise in non-linear processing. This paper introduces an interpolation technique which efficiently reduces image impulsive noise by means of an interpolant obtained through radial base functions. These functions have been used in a research project designed to develop a system for the recovery of images of Amazonian aquatic resources. This technique starts with the tagging by interpolation of noisy image pixels. Thus, a value of reconstruction for the noisy pixels is generated using neighboring pixels. The results obtained with this technique have proved comparable and often better than those obtained with previously known techniques. According to results analysis, this technique can be successfully applied on images with high noise levels. The results are low error in noisy pixel reconstruction and better visual quality.

  6. Sirenomelia with radial dysplasia.

    Science.gov (United States)

    Kulkarni, M L; Abdul Manaf, K M; Prasannakumar, D G; Kulkarni, Preethi M

    2004-05-01

    Sirenomelia is a rare anomaly usually associated with other multiple malformations. In this communication the authors report a case of sirenomelia associated with multiple malformations, which include radial hypoplasia also. Though several theories have been proposed regarding the etiology of multiple malformation syndromes in the past, the recent theory of primary developmental defect during blastogenesis holds good in this case.

  7. 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)

  8. 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

  9. Comparison of Deterministic and Probabilistic Radial Distribution Systems Load Flow

    Science.gov (United States)

    Gupta, Atma Ram; Kumar, Ashwani

    2017-12-01

    Distribution system network today is facing the challenge of meeting increased load demands from the industrial, commercial and residential sectors. The pattern of load is highly dependent on consumer behavior and temporal factors such as season of the year, day of the week or time of the day. For deterministic radial distribution load flow studies load is taken as constant. But, load varies continually with a high degree of uncertainty. So, there is a need to model probable realistic load. Monte-Carlo Simulation is used to model the probable realistic load by generating random values of active and reactive power load from the mean and standard deviation of the load and for solving a Deterministic Radial Load Flow with these values. The probabilistic solution is reconstructed from deterministic data obtained for each simulation. The main contribution of the work is: Finding impact of probable realistic ZIP load modeling on balanced radial distribution load flow. Finding impact of probable realistic ZIP load modeling on unbalanced radial distribution load flow. Compare the voltage profile and losses with probable realistic ZIP load modeling for balanced and unbalanced radial distribution load flow.

  10. The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

  11. Introducing radiality constraints in capacitated location-routing problems

    Directory of Open Access Journals (Sweden)

    Eliana Mirledy Toro Ocampo

    2017-03-01

    Full Text Available In this paper, we introduce a unified mathematical formulation for the Capacitated Vehicle Routing Problem (CVRP and for the Capacitated Location Routing Problem (CLRP, adopting radiality constraints in order to guarantee valid routes and eliminate subtours. This idea is inspired by formulations already employed in electric power distribution networks, which requires a radial topology in its operation. The results show that the proposed formulation greatly improves the convergence of the solver.

  12. 10 CFR 2.1000 - Scope of subpart J.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Scope of subpart J. 2.1000 Section 2.1000 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Procedures... Geologic Repository § 2.1000 Scope of subpart J. The rules in this subpart, together with the rules in...

  13. Analysis on Coupled Vibration of a Radially Polarized Piezoelectric Cylindrical Transducer

    Directory of Open Access Journals (Sweden)

    Jie Xu

    2017-12-01

    Full Text Available Coupled vibration of a radially polarized piezoelectric cylindrical transducer is analyzed with the mechanical coupling coefficient method. The method has been utilized to analyze the metal cylindrical transducer and the axially polarized piezoelectric cylindrical transducer. In this method, the mechanical coupling coefficient is introduced and defined as the stress ratio in different directions. Coupled vibration of the cylindrical transducer is regarded as the interaction of the plane radial vibration of a ring and the longitudinal vibration of a tube. For the radially polarized piezoelectric cylindrical transducer, the radial and longitudinal electric admittances as functions of mechanical coupling coefficients and angular frequencies are derived, respectively. The resonance frequency equations are obtained. The dependence of resonance frequency and mechanical coupling coefficient on aspect ratio is studied. Vibrational distributions on the surfaces of the cylindrical transducer are presented with experimental measurement. On the support of experiments, this work is verified and provides a theoretical foundation for the analysis and design of the radially polarized piezoelectric cylindrical transducer.

  14. Turbulence in tokamak plasmas. Effect of a radial electric field shear; Turbulence dans les plasmas de tokamaks. Effet d`un cisaillement de champ electrique radial

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs.

  15. Condition of damping of anomalous radial transport, determined by ordered convective electron dynamics

    International Nuclear Information System (INIS)

    Maslov, V.I.; Barchuk, S.V.; Lapshin, V.I.; Volkov, E.D.; Melentsov, Yu.V.

    2006-01-01

    It is shown, that at development of instability due to a radial gradient of density in the crossed electric and magnetic fields in nuclear fusion installations ordering convective cells can be excited. It provides anomalous particle transport. The spatial structures of these convective cells have been constructed. The radial dimensions of these convective cells depend on their amplitudes and on a radial gradient of density. The convective-diffusion equation for radial dynamics of the electrons has been derived. At the certain value of the universal controlling parameter, the convective cell excitation and the anomalous radial transport are suppressed. (author)

  16. 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)

  17. 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.

  18. Plasma Signatures of Radial Field Power Dropouts

    International Nuclear Information System (INIS)

    Lucek, E.A.; Horbury, T.S.; Balogh, A.; McComas, D.J.

    1998-01-01

    A class of small scale structures, with a near-radial magnetic field and a drop in magnetic field fluctuation power, have recently been identified in the polar solar wind. An earlier study of 24 events, each lasting for 6 hours or more, identified no clear plasma signature. In an extension of that work, radial intervals lasting for 4 hours or more (89 in total), have been used to search for a statistically significant plasma signature. It was found that, despite considerable variations between intervals, there was a small but significant drop, on average, in plasma temperature, density and β during these events

  19. Reble, a radially converging electron beam accelerator

    International Nuclear Information System (INIS)

    Ramirez, J.J.; Prestwich, K.R.

    1976-01-01

    The Reble accelerator at Sandia Laboratories is described. This accelerator was developed to provide an experimental source for studying the relevant diode physics, beam propagation, beam energy deposition in a gas using a radially converging e-beam. The nominal parameters for Reble are 1 MV, 200 kA, 20 ns e-beam pulse. The anode and cathode are concentric cylinders with the anode as the inner cylinder. The radial beam can be propagated through the thin foil anode into the laser gas volume. The design and performance of the various components of the accelerator are presented

  20. 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)