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Sample records for pwr primary system

  1. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

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    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Sun Jin [Hanyang Univ., Seoul (Korea, Republic of)

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author).

  2. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

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    Tanner, V.; Rosenberg, R. [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  3. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

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    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  4. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

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    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  5. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

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    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  6. Fracture mechanics evaluation for at typical PWR primary coolant pipe

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    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  7. Tribological study of hard coatings without cobalt intended to isolation components of PWR primary cooling system; Etude tribologique de revetements durs sans cobalt destines aux organes d`isolement du circuit primaire des REP

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    Cachon, L.

    1995-10-18

    The objective is to qualify coatings without cobalt to replace ``Stellites`` coatings in isolation valves of PWR primary cooling system, as Co is activated when passing in the reactor core and contaminated the cooling loop. Three families of coatings were tested: PVD thin films from 1 to 8 {mu}m monolayers of Cr/C{sub x} with x varying between 1.6 and 9.5 at% or multilayers of pure chromium and Cr/C{sub 1.6} at%, coatings with a thickness between 100 and 200 {mu}m of cermets NiCr{sub y} (y varying from 5 to 35 at%) matrix binding chromium or tungsten carbides, and thick coatings 2 mm thickness of cermets Nitronic 60 or Inconel 625 matrix binding 10, 20 or 30% titanium or niobium carbides. Stellite 6 (2 mm) is the reference coating for tribology. Coatings were qualified and selected by thermal shocks, corrosion and plane friction. The thin film and the thick families were disqualified by their destruction or by their high friction coefficient. Then coatings between 100 and 200 {mu}m were used in a valve mock-up working in PWR primary cooling system pressure and temperature conditions. Tests show that these coatings have better wear or tightness performances than stellite 6, except for a slightly higher friction coefficient. (A.B.).

  8. Conceptual study on advanced PWR system

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    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  9. Seismic qualification of PWR plant auxiliary feedwater systems

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

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14.

  10. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

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    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  11. Investigation of Burst Pressures in PWR Primary Pressure Boundary Components

    Directory of Open Access Journals (Sweden)

    Ihn Namgung

    2016-02-01

    Full Text Available In a reactor coolant system of a nuclear power plant (NPP, an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

  12. Application of LBB to high energy piping systems in operating PWR

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    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  13. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

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    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  14. Study for highly functional resin (macroporous resin) superior in removing micro particles in PWR primary circuit: on-site test

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    Itou, A.; Kondo, K.; Kouzuma, Y., E-mail: ayumu_itou@kyuden.co.jp [Kyusyu Electric Power Co., Inc., Minami-ku, Fukuoka (Japan); Umehara, R.; Shimizu, Y., E-mail: Ruyji_Umehara@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe (Japan); Kogawa, N.; Nagamine, K., E-mail: nkogawa@ndc.hq.mhi.co.jp [Nuclear Development Corp., Tokaimura, Ibaraki (Japan)

    2010-07-01

    In Japanese PWR plants, efforts to remove particulate constituents containing radioactive cobalt which provides a source of radiation exposure, are needed. Performance evaluation study was conducted for macroporous resin which was said to possess excellent performance in removing particulate constituents and whose practical accomplishment at plants in USA was reported to be good. As one of the means for radiation exposure reduction in PWR, a study for application of crud removing resin to actual plant was executed by laboratory experiments using simulated crud (Fe{sub 3}O{sub 4} particle). In this study, following two mechanisms were demonstrated as the particle capturing mechanism of macroporous resin; physical trapping by fine pores on resin surface; electrical adsorption onto resin surface. In addition, in parallel to the study for application of macroporous resin to actual PWR plant, on-site study was planned to investigate the primary system water chemistry during various stages of actual plant operation and to research performance of particle capturing in detail. As the on-site study, column experiments, there water was let pass through the column, were planned for various operation stage (startup period, power operation period and shutdown period). A kind of conventional gel-type resin and three kinds of macroporous resin were examined for onsite tests. As to particulate capturing, basic knowledge regarding capturing efficiency and influence of water chemistry on capturing performance were ordered. Capturing performance of each resin tested became clear and was ordered by comparison. Effectiveness of macroporous resin with regard to crud removal in primary coolant was confirmed. (author)

  15. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

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    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  16. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

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    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  17. Surface Oxidation Phenomena of Ni-Based Alloy 600 in PWR Primary Water Conditions

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    Lim, Yun Soo; Hwang, Seong Sik; Kim, Sung Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    There is, nevertheless, growing evidence in support for the internal oxidation model by Scot, in which grain boundary oxidation is responsible for embrittlement and cracking. Grain boundaries can act as an enhanced diffusion path for oxidation, and grain boundary oxidation can be regarded as a precursor for crack initiation. Oxidation of the grain boundary in almost all nickel-based alloys exposed to primary water is known to be detrimental for grin boundary cohesion. Panter et al. showed that the crack initiation time is strongly reduced when the specimens are pre-exposed in a simulated PWR environment in the absence of applied stress. The changes of the grain boundary structure and chemistry owing to oxygen penetration can increase the sensitivity to PWSCC under a load since grain boundary oxidization significantly weakens the grain boundary strength. Most of the important experimental results obtained are believed to correlate with the oxidation penetration into the material. A spinel structure was detected by XRD in the oxide layers. Several different types of oxide scales were found by SEM examination on the corroded surface of Alloy 600 after an immersion test in the primary water environments. Surface grain boundaries were oxidized by oxygen penetration into the matrix through grain boundaries. Grain boundary oxidization is thought to be the main reason for intergranular cracking in this alloy in a primary water environment of a PWR.

  18. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  19. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    Institute of Scientific and Technical Information of China (English)

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  20. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

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    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter [NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany); Bertholdt, Horst-Otto [NCT Consulting, Leonhardstrasse 16-18, 90443 Nuernberg (Germany); Adams, Andreas; Impertro, Michael; Roesch, Josef [RWE Power, 68643 Biblis (Germany)

    2013-07-01

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  1. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

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    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  2. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

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    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  3. Effect of water chemistry on environmentally assisted cracking of alloy 600 in simulated primary side PWR environments

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    Koenig, M. [Studsvik Nuvlear (Sweden); Lidar, P. [GSE Power Systems (Sweden); Engstroem, J. [Ringhals NPP (Sweden); Gott, K. [SKI Sweden (Sweden)

    2002-07-01

    Environmental aspects of crack growth due to intergranular stress corrosion cracking (IGSCC) of Alloy 600 in simulated primary side PWR environments have been studied. The purpose of the study was to quantify the effects of the water chemistry (Li, B and H{sub 2} concentrations, and the pH-value by adding KOH) on the crack growth rate, da/dt. 12.5 mm thick compact tension (CT) specimens were used for testing at a constant maximum stress intensity factor in the range of 26-32 MPa{open_square}m. The crack growth was continuously monitored using a direct current potential drop system. Intergranular crack growth due to IGSCC was dominant in the specimens, although there were also small fractions of transgranular cracking. Multivariate analysis was used on the results from the present work together with results from previous tests on the same material. Temperature and the stress intensity were also included as factors in the analysis. A partial least squares regression was developed and interaction effects between the factors were found to affect the crack growth rate. The Partial Least Square regression predicts the observed crack growth rates reasonably well. (authors)

  4. SCC crack growth rate of cold-worked austenitic stainless steels in PWR primary water conditions

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    Guerre, C.; Raquet, O.; Herms, E. [Commissariat a l' Energie Atomique (CEA), DEN/DPC/SCCME/LECA, Gif-sur-Yvette Cedex (France); Marie, S. [Commissariat a l' Energie Atomique (CEA), DEN/DM2S/SEMT/LISN, Gif-sur-Yvette Cedex (France); Le Calvar, M. [Inst. for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS, Fontenay-aux-Roses Cedex (France)

    2007-07-01

    Stress corrosion cracking (SCC) of stainless steels (SS) is a significant cause of failure in the pressurized water reactors (PWR). Most of the reported case history failures of SS in PWR can be attributed to pollutants (chloride, sulphate) and / or locally oxygenated environments, even to sensitisation of the SS. However, some failures have been attributed to heavy cold work (CW) of SS. In laboratory tests, SCC initiation of cold-worked SS has been obtained using slow strain rate tests (SSRT) in nominal PWR environment. This paper describes constant load and cyclic crack growth rate (CGR) tests on cold-worked SS, on CT specimens. 304L and 316L have been tested with a CW up to 60 %. CW 316L is more prone to cracking than 304L. Over 30 % of CW, 316L is susceptible to crack propagation under constant load. CW is the main controlling parameter for cracking. (author))

  5. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Energy Technology Data Exchange (ETDEWEB)

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  6. The continued development of the MFM suite and its practical application on a PWR system

    DEFF Research Database (Denmark)

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  7. Characterization of Oxide Layer with Precipitates of HANA-6 Exposed in Simulated PWR Primary Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hun; Lim, Jea Young; Lee, Sung Yong; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO NF, Daejeon (Korea, Republic of)

    2016-10-15

    The delayed oxidation behaviors of β-Nb ppts and their amorphization behaviors in HANA-6 and other Zr-base alloys have been frequently reported. On the other hand, although Zr(Nb,Fe)2 ppts could be formed in the HANA-6 alloy due to Fe impurities contained in Zrsponge, the oxidation behavior of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy has not been fully understood. In this study, oxide characteristics of HANA-6 corroded in simulated PWR environment for 165 and 315 days were investigated. And, oxidation behaviors of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy were investigated by TEM with EDS techniques. The superior corrosion property of HANA-6 has been confirmed through corrosion test in simulated PWR water for 387 days. By using TEM/EDS technique, the oxide characteristics with presence of β- Nb (or β-enriched), and ZrNbFe (possibly Zr(Nb,Fe){sub 2}) ppts have been characterized as follows. 1. Delayed oxidation behaviors of β-Nb and Zr(Nb,Fe){sub 2} ppts and their amorphization due to oxidation were observed from TEM/EDS analyses. 2. The oxide layers having crystallite and partially amorphous ppts were slightly increased with increasing corrosion test time from 165 days to 315 days. 3. In outer oxide layer, Fe in Zr(Nb,Fe){sub 2} ppt was depleted and dissolved to outer layer of ppt and bulk oxide layer.

  8. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  9. Aging assessment of PWR (Pressurized Water Reactor) Auxiliary Feedwater Systems

    Energy Technology Data Exchange (ETDEWEB)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab.

  10. Plant systems/components modularization study. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1977-07-01

    The final results are summarized of a Plant Systems/Components Modularization Study based on Stone and Webster's Pressurized Water Reactor Reference Design. The program has been modified to include evaluation of the most promising areas for modular consideration based on the level of the Sundesert Project engineering design completion and the feasibility of their incorporation into the plant construction effort.

  11. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  12. Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Thermal-hydraulic characteristic investigation on passive residual heat removal system(PRHRS)of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features.A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,a correlation of two-phase natural circulation was obtained,and relative errors of 95% test data were less than±16%.There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink,and its correlation of two-phase natural circulation system has been obtained.The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

  13. Effect of aging on the PWR Chemical and Volume Control System

    Energy Technology Data Exchange (ETDEWEB)

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-06-01

    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  14. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    Science.gov (United States)

    Chen, Junjie; Xiao, Qian; Lu, Zhanpeng; Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan

    2017-06-01

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters.

  15. Visualized Research on Primary Loop Simulation for PWR Nuclear Power Plant%压水堆核电厂一回路仿真可视化研究

    Institute of Scientific and Technical Information of China (English)

    肖瑶; 巫英伟; 苏光辉; 秋穗正

    2013-01-01

    In this study the main equipments and the primary loop of PWR nuclear power plant (NPP) were analyzed in detail.The model of point neutron dynamics,steam generator model with two-phase drift-flux governing equations,3-zone nonequilibrium pressurizer model and 4-quadrant main pump performance model were established.Based on the above models,a NPP simulation program was developed by using mixed programming with FORTRAN90 and Visual C++.The simulation program is of capability to achieve visualized simulation for the main equipments in primary loop and entire system of PWR nuclear power plant.It provides not only the visualized functions of real-time plotting,zooming,etc.,but also the output of numerical results with standard picture and/or text formatting files.Besides,the program was validated by comparing the calculation results of the program developed by authors and those of RELAP5/MOD3.0.%对压水堆核电厂一回路系统及主要设备进行了详细分析,建立了点堆中子动力学模型、两相漂移流蒸汽发生器模型、三区不平衡稳压器模型和主循环泵四象限特性模型,并以此为基础使用FORTRAN90语言和Visual C++语言通过混合编程的方法开发了核电厂仿真分析程序,实现了对压水堆核电厂一回路主要设备及全系统的可视化仿真计算.软件提供实时绘图、缩放等可视化功能,还提供了数据结果的标准图片格式和标准文本格式输出.通过将程序的计算结果与RELAP5/MOD3.0计算结果进行比较,对程序的可靠性进行了验证.

  16. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  17. Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu, E-mail: wangjianqiu@imr.ac.cn; Han, En-Hou; Ke, Wei

    2015-05-15

    Highlights: • A duplex oxide film can be formed on the Welded 308L. • Surface state has no influence on the phase composition of the oxide film. • Surface state can affect the thickness of the oxide film. • Surface state can affect the morphology of the oxide film. - Abstract: The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe{sub 3}O{sub 4}, Fe{sub 2}O{sub 3} and a small amount of NiFe{sub 2}O{sub 4}, Ni(OH){sub 2}, Cr(OH){sub 3} and (Ni, Fe)Cr{sub 2}O{sub 4}) and a Cr-rich inner layer (FeCr{sub 2}O{sub 4} and NiCr{sub 2}O{sub 4}) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

  18. Determination of the {sup 129}I in primary coolant of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ke Chon; Park, Yong Joon; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-02-15

    Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as {sup 129}I is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for {sup 129}I as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding {sup 125}I as a radio-isotopic tracer (t1/2 = 60.14 d) to the simulation sample, in order to measure the activity concentration of {sup 129}I in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, Clform) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of {sup 129}I was examined, as was the effect of {sup 3}H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous {sup 3}H presence was found with activity concentrations of {sup 3}H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 {mu}g/mL.

  19. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  20. Stress corrosion crack initiation of alloy 600 in PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Toloczko, Mychailo B.; Olszta, Matthew J.; Bruemmer, Stephen M.

    2017-07-01

    Stress corrosion crack (SCC) initiation of three mill-annealed (MA) alloy 600 heats in simulated pressurized water reactor primary water has been investigated using constant load tests equipped with in-situ direct current potential drop (DCPD) measurement capabilities. SCC initiation times were greatly reduced by a small amount of cold work. Shallow intergranular (IG) attack and/or cracks were found on most high-energy grain boundaries intersecting the surface with only a small fraction evolving into larger cracks and IGSCC growth. Crack depth profiles were measured and related to DCPD-detected initiation response. Processes controlling the SCC initiation in MA alloy 600 are discussed. IN PRESS, CORRECTED PROOF, 05/02/2017 - mfl

  1. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  2. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  3. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  4. The Effects of Hot Bending on the Low Cycle Fatigue Behaviors of 347 SS in PWR Primary Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho-Sub; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Fatigue damage could be significant for some locations, especially the welds and bends where stress concentration is typically high. As a possible solution, a large radius hot-bending method has been suggested to eliminate some weld joints and all tight bends. However, for the hot-bending process which involves a high temperature thermal cycle, there is a concern about changes in mechanical properties including low cycle fatigue behaviors. In APR1400, Type 347 SS have been used as surge line pipes. Therefore, to verify the applicability of hot-bending on 347 SS surge line pipes, an environmental fatigue test program was initiated. In this paper, the preliminary results of the on-going test program are introduced. Also, the low cycle fatigue behaviors of 347 SS are compared with those of other grade of stainless steels. The effects of hot bending on the low cycle fatigue behavior of 347 SS were quantitatively evaluated. The fatigue life was compared with the estimated values per NUREG 6909 rev. 1. There are no distinct differences between NUREG 6909 and LCF tests. According to fractography and cross section analysis in progress, basically, the reduction of LCF life of 347 SS in PWR water was caused by operation of HIC mechanism. The cyclic stress responses shows that there is no secondary hardening in 330 .deg.C air and PWR water.

  5. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  6. Materials Reliability Program: Environmental Fatigue Testing of Type 304L Stainless Steel U-Bends in Simulated PWR Primary Water (MRP-137)

    Energy Technology Data Exchange (ETDEWEB)

    R.Kilian

    2004-12-01

    Laboratory data generated in the past decade indicate a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. In recent comprehensive review of laboratory, component and structural test data performed through the EPRI Materials Reliability Program, flow rate was identified as a critical variable that was generally not considered in laboratory studies but applicable in plant operating environments. Available data for carbon/low-alloy steel piping components suggest that high flow is beneficial regarding the effects of a reactor water environment. Similar information is lacking for stainless steel piping materials. This report documents progress made to date in an extensive testing program underway to evaluate the effects of flow rate on the corrosion fatigue of 304L stainless steel under simulated PWR primary water environmental conditions.

  7. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    Energy Technology Data Exchange (ETDEWEB)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S. [Electricite de France (EDF), 78 - Chatou (France). Lab. National d`Hydraulique

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author) 12 refs.

  8. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yichun, E-mail: ycwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: 807001564@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: 1056303902@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: 1032133755@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: npic_wu@126.com [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: zwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China)

    2016-05-15

    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  9. Effects of cold work and stress on oxidation and SCC behavior of stainless steels in PWR primary water environments

    Energy Technology Data Exchange (ETDEWEB)

    Shoji, T.; Sakaguchi, K.; Lu, Z. [Fracture and Reliability Research Institute, Tohoku University, Sendai City 980-8579 (Japan); Hirano, S.; Hasegawa, Y. [Kansai Electric Power Co (Japan); Kobayashi, T.; Fujimoto, K.; Nomura, Y. [Mitsubishi Heavy Industries (Japan)

    2011-07-01

    Intergranular stress corrosion cracking (SCC) samples taken from a weld HAZ of 316 stainless steel welded to a low alloy steel of steam generator nozzle with nickel base alloy 82 in Mihama Unit 2 PWR plant were analyzed by extensive metallographic observation, micro-Raman spectroscopy, TEM analysis of stainless steel material, oxide morphology, compositional profiles as well as their crystal structures. The crack growth history during the plant operation is discussed in connection to a residual stress distribution at HAZ and distribution of oxides on/in the cracks. Possible time dependence of crack growth rate with crack growth in components was proposed based upon the evidences observed about oxides. The importance of surface integrity assessment in SCC initiation and propagation is emphasized from a point of view of oxidation localization which can be promoted by strain (dislocation density), straining and stress, which play a crucial role in oxidation due to accelerated mass transfer in oxides as well as underlying metallic materials. Especially, preferential oxidation along slip bands suggests that oxygen diffusion in such a region with a high dislocation density is faster than the other region. This fact implies that grain boundary can also be a preferential path of oxidation as has been observed by TEM, TOFSIMS and 3D-APT. This localization of oxidation and acceleration is discussed based upon an analysis of profile development at a stressed oxide/metal interface. The effects of environmental parameters, temperature, loading mode, and rolling procedures on SCC of stainless steels in simulated PWR environments were investigated by laboratory tests. Strong interactions among grain boundary structure, environmental parameters and interfacial oxidation kinetics, and SCC behavior are observed

  10. The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

  11. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  12. Practical Application of the MFM Suite on a PWR System: Modelling and Reasoning on Causes and Consequences of Process Anomalies

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Thunem, Harald P - J; Lind, Morten

    2014-01-01

    Multilevel Flow Modelling (MFM) is a functional modelling methodology which applies means - end and parts - whole decomposition and aggregation techniques to handle the complexity of engineering systems. It has been adopted in several case studies to model the process goal and functions of PWR...... is equipped with an MFM Model Editing Interface to facilitate the modelling process and MFM model analysis modules to run diag nosis and prognosis analyses based on developed models. New features of the MFM Suite also include making corresponding process diagram for the plant being modelled with MFM...... and linking the MFM model to its process components. The purpose of this report is to make a comprehensive demonstration of how to use the MFM Suite to develop MFM models and run causal reasoning for abnormal situations. This report will explain the capability of representing process and operational knowledge...

  13. The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, R B; Hua, F H

    2004-07-12

    Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attempt is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.

  14. TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-08-01

    Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

  15. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  16. Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment

    Directory of Open Access Journals (Sweden)

    Sang-Won Lee

    2014-01-01

    Full Text Available After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to 10-3 considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.

  17. Seismic Spectrum Analysis of Advanced PWR Primary Loop and Parameter Sensitivity Study%先进压水堆一回路地震反应谱分析及参数敏感性研究

    Institute of Scientific and Technical Information of China (English)

    段蓉; 佟立丽; 曹学武

    2014-01-01

    The primary loop system of pressurized water reactor (PWR ) consists of reactor pressure vessel (RPV) ,steam generator (SG) ,reactor coolant pump (RCP) , pressurizer ,surge line and other crucial components .T he seismic response of each com-ponent is closely related to the structure of the w hole system .From a system perspec-tive ,the primary loop system of the advanced passive pressurized water reactor was studied .With ANSYS software ,the three-dimensional finite element model was built to perform modal analysis .Based on the results of modeling ,the seismic spectrum analysis with three orthotropic directions on the primary loop system was conducted to obtain the stress and displacement response .T hen sensitivity analysis of parameters ,such as spec-trum input angle and stiffness of supports was performed ,giving guidance on further design and analysis .Moreover ,direct integration method was used to get time-history response ,which was compared with spectrum simulation results .The displacement and acceleration input for seismic analysis of specific components were offered .Besides , compared with three-dimensional finite element model ,the lumped mass model of the primary loop system was also built to conduct seismic analysis ,giving the advantage and necessity of three-dimensional modeling .The study provides technical support for the structure analysis of key equipments of advanced PWR primary loop .%压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用 ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数

  18. Physics of hydride fueled PWR

    Science.gov (United States)

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  19. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  20. 压水堆主管道上充管嘴弹塑性应力分析%Elastoplastic Stress Analysis for Charging Nozzle of PWR Primary Piping

    Institute of Scientific and Technical Information of China (English)

    艾红雷; 郑斌; 卢喜丰; 王新军

    2016-01-01

    压水堆主管道上充管嘴在核电厂运行期间需经受严厉的冷热流体交互流动产生的循环热载荷,将对上充管嘴的结构完整性产生重要的影响.上充管嘴弹性应力分析证明了结构不会出现弹性失稳、塑性失稳以及疲劳破坏等现象,但部分分析截面一次加二次应力强度范围超过了规范限值.针对弹性分析部分结果不满足规范限值的情况,对上充管嘴进行了循环弹塑性分析,结果表明,上充管嘴结构在循环载荷作用下出现了明显的塑性安定现象,并且经历所分析的循环载荷后,其结构的累积应变不会对结构抗塑性失稳能力和抗疲劳破坏能力产生显著的影响.上充管嘴抗快断分析表明,其结构具备良好的抗快断性能.%The severe cyclical thermal loadings which generated by the interaction flow of hot and cold flu-id are subjected to the charging nozzle of PWR primary piping during the operation of nuclear power plants.The loadings have significant impact on the structural integrity.The types of damage( elastic insta-bility,plastic instability,fatigue and so on) do not occur for the nozzle is proved by elastic analysis.For some analysis sections,the range of primary plus secondary stresses cannot be validated and the cyclical elastoplastic analysis is used.The results show that the behavior of plastic accommodation occurred under the cyclical loadings and the cumulative strain induced by repeated loads do not diminish the capacity to prevent the damage of plastic instability and fatigue.The results of resistance to fast fracture analysis show that the nozzle has a good ability of resistance to fast fracture.

  1. Stress corrosion cracking of Ni-based alloys in PWR primary water. Component surface control; Corrosion sous contrainte des alliages a base nickel en milieu primaire des reacteurs a eaux pressurisee. Maitrise de la surface des composants

    Energy Technology Data Exchange (ETDEWEB)

    Foucault, M. [AREVA, Centre Technique Framatome ANP, Dept. Corrosion Chimie, 71 - Le Creusot (France)

    2004-06-01

    In the PWR plant primary circuit, FRAMATOME-ANP uses several nickel-base alloys or austenitic stainless steels for the manufacture of safety components. The experience feedback of the last twenty years allows us to point out the major role played by the surface state of the components in their life duration. In this paper, we present two examples of problems encountered and solved by a surface study and the definition and implementation of a process for the surface control of the repair components. Then, we propose some ideas about the present needs in terms of analysis methods to improve the surface knowledge and the control of the manufactured components. (author)

  2. Effects of dissolved hydrogen on general corrosion behavior and oxide films of alloy 690TT in PWR primary water

    Science.gov (United States)

    Jeon, Soon-Hyeok; Lee, Eun-Hee; Hur, Do Haeng

    2017-03-01

    The effect of dissolved hydrogen (DH) on the general corrosion behavior and oxide films of Alloy 690TT is investigated in simulated primary water at 330 °C. With increasing DH, the structure of oxide film significantly changed and the corrosion rate decreased. At DH = 5 cm3/kg H2O, the oxide layer was thick, and consisted of outer Ni oxide layer and inner Cr2O3 layer. Under the conditions of DH = 35 and 100 cm3/kg H2O, the oxide films grew thinner and composed of outer polyhedral spinel oxide particles such as NiCr2O4 or NiCrFeO4 and an intermediate metallic Ni-rich layer, with inner Cr2O3 layer. The general corrosion rate significantly decreased by about 72% as DH concentration increased from 5 to 35 cm3/kg H2O. In the range of 35-65 cm3/kg H2O, the corrosion rate slightly decreased with increasing DH concentration. However, no further changes were observed in the range of 65-100 cm3/kg H2O.

  3. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  4. Effects of aging in containment spray injection system of PWR reactor containment; Efeitos do envelhecimento no sistema de injecao de borrifo da contencao de reatores a agua pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: diogosb@outlook.com, E-mail: deise_dy@hotmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems.

  5. Qualification of NEXUS/ANC Nuclear Design System for PWR Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Mayhue, Larry; Milanova, Radka; Huria, Harish; Zhang, Baocheng; Franceschini, Fausto; Ouisloumen, Mohamed [Westinghouse Electric Company, Pittsburgh, PA (United States); Mueller, Erwin; Forslun Guimaraes, Petri [Westinghouse Electric Company, Vaesteraas (Sweden)

    2008-07-01

    NEXUS is a new cross section and nuclear data generation system for core simulators developed by Westinghouse. This system generates once-through, full temperature range nuclear data for both PWRs and BWRs. The system has been implemented for PWRs in the NEXUS/ANC code system. A brief description of the methodology and the codes comprising this system is presented. The qualification for NEXUS/ANC has been completed and a summary of some of the results is presented for 10 plants and 45 cycles of operation. These results include startup data and at-power axial offset performance. Results for low temperature calculations are also presented. The NEXUS/ANC system includes new methodology to cover the operation of AP1000 plants including a new pin power recovery method and a method to capture the effects of control rod depletion. A brief summary of these methods is also presented. (authors)

  6. Failure Modes and Effects Analysis (FMEA) of the Residual Heat Removal System. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Eggleston, F T

    1976-01-01

    The Residual Heat Removal System (RHRS) transfer heat from the Reactor Coolant System (RCS) to the reactor plant Component Cooling System (CCS) to reduce the temperature of the RCS at a controlled rate during the second part of normal plant cooldown and maintains the desired temperature until the plant is restarted. By the use of an analytic tool, the Failure Modes and Effects Analysis, it is shown that the RHRS, because of its redundant two train design, is able to accommodate any credible component single failure with the only effect being an extension in the required cooldown time, thus demonstrating the reliability of the RHRS to perform its intended function.

  7. Initial guidance on digraph-matrix analysis for systems interaction studies. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Alesso, H.P.; Sacks, I.J.; Smith, C.F.

    1983-03-01

    This report describes the digraph-matrix analysis for systems structural analysis. The method is useful to analysts that are searching for both single failures and paired failures that disable systems. The digraph-matrix analysis can assure the analyst that the independent functioning of a safety system is not jeopardized by design features that cause faults to be dependent. The digraph-matrix analysis facilitates the discovery and the quantification of component reachability. The guidance is sufficiently specific that the reader can make direct application. Because a systems interaction analysis of an LWR is expensive, the resource efficiency of a candidate method is important to the staff. A demonstration of the digraph-matrix analysis is part of the staffs efforts to provide a measure of its resource efficiency. Additionally, there are features within the digraph-matrix analysis itself that might be modified to enhance resource efficiency.

  8. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

  9. PWR hybrid computer model for assessing the safety implications of control systems

    Energy Technology Data Exchange (ETDEWEB)

    Smith, O L; Renier, J P; Difilippo, F C; Clapp, N E; Sozer, A; Booth, R S; Craddick, W G; Morris, D G

    1986-03-01

    The ORNL study of safety-related aspects of nuclear power plant control systems consists of two interrelated tasks: (1) failure mode and effects analysis (FMEA) that identified single and multiple component failures that might lead to significant plant upsets and (2) computer models that used these failures as initial conditions and traced the dynamic impact on the control system and remainder of the plant. This report describes the simulation of Oconee Unit 1, the first plant analyzed. A first-principles, best-estimate model was developed and implemented on a hybrid computer consisting of AD-4 analog and PDP-10 digital machines. Controls were placed primarily on the analog to use its interactive capability to simulate operator action. 48 refs., 138 figs., 15 tabs.

  10. Experimental Research on Passive Residual Heat Removal System of Chinese Advanced PWR

    Directory of Open Access Journals (Sweden)

    Zhuo Wenbin

    2014-01-01

    Full Text Available Passive residual heat removal system (PRHRS for the secondary loop is one of the important features for Chinese advance pressurized water reactor (CAPWR. To prove the safety characteristics of CAPWR, serials of experiments have been done on special designed PRHRS test facility in the former stage. The test facility was built up following the scaling laws to preserve the similarity to CAPWR. A total of more than 300 tests have been performed on the test facility, including 90% steady state cases and 10% transient cases. A semiempirical model was generated for passive heat removal functions based on the experimental results of steady state cases. The dynamic capability characteristics and reliability of passive safety system for CAPWR were evidently proved by transient cases. A new simulation code, MISAP2.0, has been developed and calibrated by experimental results. It will be applied in future design evaluation and optimization works.

  11. Development of a Multi-Group Neutron Cross Section Library Generation System for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Hong, Ser Gi; Song, Jae Seung; Lee, Kyung Hoon; Cho, Jin Young; Kim, Ha Yong; Koo, Bon Seung; Shim, Hyung Jin; Park, Sang Yoon

    2008-10-15

    This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared.

  12. Risk Analysis of an interfacing system LOCA in a generic Westinghouse PWR

    OpenAIRE

    Favre, Jean-Baptiste

    2014-01-01

    This project has been developed during my internship in the field of Probabilistic Safety Assessment (PSA) in the offices of Westinghouse. These are in the enclosure of Vandellòs’ Nuclear Plant in Hospitalet de l'Infant. The goal of my internship was the modelling and computation of the frequency that an Interfacing System Loss of Coolant Accident (ISLOCA) occurs in the nuclear power plants of Vandellòs and Ascó. In order to achieve this goal, I applied a method to calculate...

  13. Advanced water processing system (AWPS), including advanced filtration system (AFS) and advanced ion selective system (AISS) for improved utility (PWR/BWR) water processing performance

    Energy Technology Data Exchange (ETDEWEB)

    Denton, Mark S. [ATG, Inc.(United States); Vance, Jene N. [V and V, Inc. (United States)

    1999-07-01

    The advanced water processing system (AWPS) has the potential for wide spread success on a worldwide scale in both PWR and BWRs. The AWPS incorporates the advanced features (patent pending) of advanced filtration and advanced ion selective technologies (patented). Typical problems encountered in current filtration systems include: (1) poor effluent quality, (2) short run lengths on filters, (3) frequent filter change-outs/backwashes, (4) large waste volumes, and (5) failed filter cartridges. The advanced filtration system (AFS) features reduced waste production per million gallons of water processed, cleaner water for recycle or release to the environment, filter element volume 100 times less than that of competitive filters, and a far lower capital cost compared to systems with similar performance. The AWPS should be of interest to plants that are upgrading, or to new plants to lower both their capital and operating costs, as well as total curie discharge levels. In addition, the AWPS will function in non-nuclear, as well as nuclear, applications of water purification, specially where pre coat filtration/ion exchange or reverse osmosis (RO) is being applied to process water with high concentrations of colloidal contaminants. Pilot testing has been successfully completed in the U. S. at the Byron (PWR), LaSalle (BWR), and Dresden(BWR) nuclear plants for Commonwealth Edison, and the Bruce several spent filters in a High Integrated Container these bench- and pilot-scale demonstrations will be presented herein. Full-scale designs or systems have been shipped to these locations. In all cases, the testing demonstrated: (1) longer run lengths (300,000 gallons between backwashes--a 100 fold improvement), (2) recoverability of cartridge filters after backwash (cartridge lives of approximately 6 months to a year--a 5 to 10 fold improvement in filter life), (3) large removal efficiencies for colloidal particles (reduced discharge curies), and (4) reduced waste volumes

  14. 400-MWe Consolidated Nuclear Steam System (CNSS). 1200-MWt Phase 2A interim studies. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

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

  15. Clay Generic Disposal System Model - Sensitivity Analysis for 32 PWR Assembly Canisters (+2 associated model files).

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Edgar [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    The Used Fuel Disposition Campaign (UFDC), as part of the DOE Office of Nuclear Energy’s (DOE-NE) Fuel Cycle Technology program (FCT) is investigating the disposal of high level radioactive waste (HLW) and spent nuclear fuela (SNF) in a variety of geologic media. The feasibility of disposing SNF and HLW in clay media has been investigated and has been shown to be promising [Ref. 1]. In addition the disposal of these wastes in clay media is being investigated in Belgium, France, and Switzerland. Thus, Argillaceous media is one of the environments being considered by UFDC. As identified by researchers at Sandia National Laboratory, potentially suitable formations that may exist in the U.S. include mudstone, clay, shale, and argillite formations [Ref. 1]. These formations encompass a broad range of material properties. In this report, reference to clay media is intended to cover the full range of material properties. This report presents the status of the development of a simulation model for evaluating the performance of generic clay media. The clay Generic Disposal System Model (GDSM) repository performance simulation tool has been developed with the flexibility to evaluate not only different properties, but different waste streams/forms and different repository designs and engineered barrier configurations/ materials that could be used to dispose of these wastes.

  16. Influence of localized deformation on A-286 austenitic stainless steel stress corrosion cracking in PWR primary water; Influence de la localisation de la deformation sur la corrosion sous contrainte de l'acier inoxydable austenitique A-286 en milieu primaire des REP

    Energy Technology Data Exchange (ETDEWEB)

    Savoie, M

    2007-01-15

    Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels is known to be a critical issue for structural components of nuclear reactor cores. The deformation of irradiated austenitic stainless steels is extremely heterogeneous and localized in deformation bands that may play a significant role in IASCC. In this study, an original approach is proposed to determine the influence of localized deformation on austenitic stainless steels SCC in simulated PWR primary water. The approach consists in (i) performing low cycle fatigue tests on austenitic stainless steel A-286 strengthened by {gamma}' precipitates Ni{sub 3}(Ti,Al) in order to shear and dissolve the precipitates in intense slip bands, leading to a localization of the deformation within and in (ii) assessing the influence of these {gamma}'-free localized deformation bands on A-286 SCC by means of comparative CERT tests performed on specimens with similar yield strength, containing or not {gamma}'-free localized deformation bands. Results show that strain localization significantly promotes A-286 SCC in simulated PWR primary water at 320 and 360 C. Moreover, A-286 is a precipitation-hardening austenitic stainless steel used for applications in light water reactors. The second objective of this work is to gain insights into the influence of heat treatment and metallurgical structure on A-286 SCC susceptibility in PWR primary water. The results obtained demonstrate a strong correlation between yield strength and SCC susceptibility of A-286 in PWR primary water at 320 and 360 C. (author)

  17. SCOR 1000: an economic and innovative conceptual design PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  18. PWR reactor vessel in-service-inspection according to RSEM

    Energy Technology Data Exchange (ETDEWEB)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  19. 基于图论的压水堆核电机组能耗定量分析模型%A General Model Based on Graph Theory for Quantitative Analysis of PWR Thermodynamic System

    Institute of Scientific and Technical Information of China (English)

    冉鹏; 王亚瑟

    2013-01-01

    Based on the analysis of the structure feature of PWR nuclear power plants, graph theory are introduced in the thermal economy analysis fields. According to the abstraction rule of the thermal system in PWR nuclear power plants, the boundary delimitation of a power plant thermal system is determined, and the thermal system of PWR nuclear power plants is expressed as the form of graph theory. A new unified rules for analyzing the thermal system are established. Combined with the first thermodynamics law and mass conservation law, weighted diagraph adjacency matrix is deducted. An example is given to illustrate the validity of the method.%在深入研究压水堆(PWR)核电机组热力系统结构特点的基础上,将图论思想引入热力系统节能分析,规定核电机组热力系统的划分原则及其基于图的表达方法,确定核电机组热力系统的有向图带权邻接矩阵填写规则.根据回热加热器系统的能量守恒定律、质量守恒定律,确定核电机组主、辅系统的有向图带权邻接矩阵表达规则以及矩阵的运算规则,推导出通用PWR核电机组热力系统的有向图带权邻接矩阵方程,并用实例验证本方法的正确性.

  20. Mitigation of stress corrosion cracking in pressurized water reactor (PWR) piping systems using the mechanical stress improvement process (MSIP{sup R)} or underwater laser beam welding

    Energy Technology Data Exchange (ETDEWEB)

    Rick, Grendys; Marc, Piccolino; Cunthia, Pezze [Westinghouse Electric Company, LLC, New York (United States); Badlani, Manu [Nu Vision Engineering, New York (United States)

    2009-04-15

    A current issue facing pressurized water reactors (PWRs) is primary water stress corrosion cracking (PWSCC) of bi metallic welds. PWSCC in a PWR requires the presence of a susceptible material, an aggressive environment and a tensile stress of significant magnitude. Reducing the potential for SCC can be accomplished by eliminating any of these three elements. In the U.S., mitigation of susceptible material in the pressurizer nozzle locations has largely been completed via the structural weld overlay (SWOL) process or NuVision Engineering's Mechanical Stress Improvement Process (MSIP{sup R)}, depending on inspectability. The next most susceptible locations in Westinghouse designed power plants are the Reactor Vessel (RV) hot leg nozzle welds. However, a full SWOL Process for RV nozzles is time consuming and has a high likelihood of in process weld repairs. Therefore, Westinghouse provides two distinctive methods to mitigate susceptible material for the RV nozzle locations depending on nozzle access and utility preference. These methods are the MSIP and the Underwater Laser Beam Welding (ULBW) process. MSIP applies a load to the outside diameter of the pipe adjacent to the weld, imposing plastic strains during compression that are not reversed after unloading, thus eliminating the tensile stress component of SCC. Recently, Westinghouse and NuVision successfully applied MSIP on all eight RV nozzles at the Salem Unit 1 power plant. Another option to mitigate SCC in RV nozzles is to place a barrier between the susceptible material and the aggressive environment. The ULBW process applies a weld inlay onto the inside pipe diameter. The deposited weld metal (Alloy 52M) is resistant to PWSCC and acts as a barrier to prevent primary water from contacting the susceptible material. This paper provides information on the approval and acceptance bases for MSIP, its recent application on RV nozzles and an update on ULBW development.

  1. Metallurgical and mechanical parameters controlling alloy 718 stress corrosion cracking resistance in PWR primary water; Facteurs metallurgiques et mecaniques controlant l'amorcage de defauts de corrosion sous contrainte dans l'alliage 718 en milieu primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Deleume, J

    2007-11-15

    Improving the performance and reliability of the fuel assemblies of the pressurized water reactors requires having a perfect knowledge of the operating margins of both the components and the materials. The choice of alloy 718 as reference material for this study is justified by the industrial will to identify the first order parameters controlling the excellent resistance of this alloy to Stress Corrosion Cracking (SCC). For this purpose, a specific slow strain rate (SSR) crack initiation test using tensile specimen with a V-shaped hump in the middle of the gauge length was developed and modeled. The selectivity of such SSR tests in simulated PWR primary water at 350 C was clearly established by characterizing the SCC resistance of nine alloy 718 thin strip heats. Regardless of their origin and in spite of a similar thermo-mechanical history, they did not exhibit the same susceptibility to SCC crack initiation. All the characterized alloy 718 heats develop oxide scale of similar nature for various exposure times to PWR primary medium in the temperature range [320 C - 360 C]. {delta} phase precipitation has no impact on alloy 718 SCC initiation behavior when exposed to PWR primary water, contrary to interstitial contents and the triggering of plastic instabilities (PLC phenomenon). (author)

  2. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  3. Effect of co-free valve on activity reduction in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  4. ALIBABA, an assistance system for the detection of confinement leaks in a PWR reactor; ALIBABA, un systeme d`aide a la detection des voies de fuites du confinement sur un reacteur a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Bedier, P.O.; Libmann, M. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

    1995-12-31

    The objective of the Crisis Technical Center (CTC) of the French Institute for Nuclear Protection and Safety (IPSN) is to estimates the consequences of a given nuclear accident on the populations and the environment. ALIBABA is a data processing tool available at the CTC and devoted to the detection of confinement leaks in 900 MWe PWR reactors using the activity values measured by the captors of the installation. The heart of this expert system is a structural and functional representation of the different components directly involved in the leak detection (isolating valves, ventilation systems, electric boards etc..). This tool can manage the availability of each component to make qualitative and quantitative balance-sheets. This paper presents the ALIBABA software, an industrial prototype realized with the SPIRAL knowledge base systems generator at the CEA Reactor Studies and Applied Mathematics Service (SERMA) and commercialized by CRIL-Ingenierie Society. It describes the techniques used for the modeling of PWR systems and for the visualization of the survey. The functionality of the man-machine interface is discussed and the means used for the validation of the software are summarized. (J.S.). 6 refs.

  5. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain; Recomendaciones del MRP-139: Inspeccion de soldaduras disimilares en Vasijas de Reactor en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-07-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  6. Design and Retrofit of Radiation Monitoring System for the PWR Nuclear Power Plant%压水堆核电厂辐射监测系统的设计与改造

    Institute of Scientific and Technical Information of China (English)

    张涛; 熊国华; 郎玉凯; 郭伟

    2011-01-01

    辐射监测系统是压水堆核电厂安全运行的重要保障,研究压水堆核电厂辐射监测系统的设计方法和原则,对于提高压水堆核电厂辐射监测系统的设计水平,减少改造风险至关重要.根据核电厂的法规和设计规范,结合大亚湾核电厂辐射监测系统的设计与改造经验,提出了压水堆核电厂辐射监测系统的一般设计原则和要求,并简要介绍了大亚湾核电厂辐射监测系统的改造措施及方法.%Radiation monitoring system is important for the PWR nuclear power plant, and the research of design methods and principles for the radiation monitoring system can greatly improve the design ability of the system for PWR nuclear power plant, and reduce the risk of system retrofit. According to the Nuclear power plant regulations and design specifications, and taking the design and retrofit experience of the radiation monitoring system in Daya Bay Nuclear Power Plant into account, the general design principles and requirements of the radiation monitoring system in the PWR nuclear power plant is proposed, and the retrofit method of the radiation monitoring system in Daya Bay Nuclear Power Plant is briefly introduced.

  7. Advanced ion exchange resins for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

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

  8. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Directory of Open Access Journals (Sweden)

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  9. Retention of PWR primary coolant trace elements by cation exchange resins during cold shutdown with oxygenation: modelling and experimental results for silver behavior; Retention des elements traces du fluide primaire des REP par les resines echangeuses de cations lors des mises en arret a froid avec oxygenation: modelisation et resultats experimentaux relatifs au comportement de l'argent

    Energy Technology Data Exchange (ETDEWEB)

    Elain, L.; Doury-Berthod, M. [CEA Saclay, INSTN, Institut National des Sciences et Techniques Nucleaires, 91 - Gif-sur-Yvette (France); Genin, J.B. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Berger, M. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France)

    2004-07-01

    In order to minimize the radiochemical impact of the corrosion products on the operation of Pressurized Water Reactors, on-line purification of the primary coolant is carried out. The purification system arranged on the Chemical and Volume Control System is made up of mechanical filters and demineralizers packed with a mixed bed of cation and anion exchange resins. This paper proposes an update on the retention of primary coolant trace elements by the cation exchange resins of the demineralizers during cold shutdowns with oxygenation. The study is first of all devoted to the description of the concentration profiles of the various cation constituents which settle in the demineralizer during purification after oxygenation. For a number of trace elements, localized enrichment zones at the Li{sup +}/Ni(Il) exchange zone are expected to appear in the column. The case of silver is afterwards discussed in detail. Thermodynamic modelling shows that the theoretical retention volume of the metallic element and its degree of enrichment in the column are dependent on the basic composition of the primary coolant and the specific characteristics of the demineralizer cation exchanger. At the Ag{sup +} ion concentration expected in the reactor coolant after oxygenation (between 10{sup -8} mol.L{sup -1} and 10{sup -6} mol.L{sup -1}), the breakthrough of silver should be near-simultaneous with that of nickel. The experimental results, obtained in the laboratory and with a 'Mini-CVCS' pilot instrumentation recently used during the cold shutdown of Tricastin Unit 2,900 MWe PWR NPP, confirm the validity of these theoretical forecasts and enable new hypotheses to be advanced for explaining silver release from a demineralizer. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

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

  11. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott; Hsu, C.

    1999-08-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  12. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    Energy Technology Data Exchange (ETDEWEB)

    A. G. Ware; C. Hsu (USNRC); C. L. Atwood; M. B. Sattison; R. S. Hartley (INEEL); V. N. Shah

    1999-02-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  13. Information system of corrosion and mechanical properties for steels used in nuclear power plants with PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, M.; Novotny, R.; Sajdl, P. [Inst. of Chemical Technology, Prague (Czech Republic). Dept. of Power Engineering

    1998-11-01

    This paper gives information about a new developed database system which contains information about chemical constitution of steels used in nuclear power plants. It enables to hold data from corrosion tests and allows to insert graphs and pictures into the form. This system is an application of MS Access. (orig.)

  14. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.H. [Duke Power Co., Seneca, SC (United States)

    1995-04-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT{trademark} Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS.

  15. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  16. 压水堆核电机组反应堆系统仿真实现%The Realization of PWR Reactor System Simulation

    Institute of Scientific and Technical Information of China (English)

    郭俊伟; 陈启卷

    2014-01-01

    本文根据压水堆的物理特性,综合运用中子动力学、温度反馈效应、中毒效应、堆芯热传递等相关数学模型,建立了适用于PC使用的压水堆核电站反应堆本体的仿真模块。然后利用Matlab/Simulink实现了该模型的实时仿真,建立了包括中子动力学仿真子模块、氙中毒效应仿真子模块、温度反馈仿真子模块和堆芯热传输仿真子模块在内的四个仿真模块。最后封装成压水堆核电站反应堆系统仿真模块。所建立的反应堆仿真模块基于点堆模型,从数值仿真结果来看,由它导出的结果令人满意。该模型可用于分析反应堆的大部分瞬态过程,解释堆内中子通量密度随时间变化的大部分特性,研究局部扰动对反应堆堆芯参数的影响。%On the basis of PWR physical characteristics and the integrated use of neutron kinetics, temperature feedback effect, poisoning effect, core heat transfer and other related mathematics model, the nuclear power station's core physical and mathematical models applicable to a PC simulation has been established. Then the real-time simulation model, which including Neutron Dynamics simulation module, Xenon Poisoning Effects simulation module, Temperature Feedback simulation module and Core Heat Transfer simulation module, is realized by the Matlab / Simulink simulation tool. Finally they are integrated into nuclear power plant Reactor System simulation module. Simulation results show that the mathematical models have high degree of accuracy. The model can be used to analyze the most transient reactor process, explain neutron flux density properties with time, and research the partial perturbation's influence to reactor parameters.

  17. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  18. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  19. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Energy Technology Data Exchange (ETDEWEB)

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  20. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  1. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  2. Integral Test Facility PKL: Experimental PWR Accident Investigation

    OpenAIRE

    2012-01-01

    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  3. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  4. PWR Cross Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  5. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  6. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  7. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  8. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  9. Preoperational test report, primary ventilation system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-11-04

    This represents a preoperational test report for Primary Ventilation Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system provides vapor space filtered venting of tanks AY101, AY102, AZ101, AZ102. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  10. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Energy Technology Data Exchange (ETDEWEB)

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  11. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  12. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  13. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  14. Neutron noise measurements on Bugey 2 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Marini, J.; Romy, D.; Spadi, J.C.; Assedo, R.; Castello, G.

    1982-01-01

    Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurements runs during the first cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

  15. A multi-agent design for a pressurized water reactor (P.W.R.) control system; Modelisation multi-agents pour la conduite d'un reacteur a eau sous pression (REP)

    Energy Technology Data Exchange (ETDEWEB)

    Aimar-Lichtenberger, M. [Paris-11 Univ., 91 - Orsay (France)

    1999-01-01

    This PhD work is in keeping with the complex industrial process control. The starting point is the analysis of control principles in a Pressurized Water Reactor (P.W.R). In order to cope with the limits of the present control procedures, a new control organisation by objectives and means is defined. This functional organisation is based on the state approach and is characterized by the parallel management of control functions to ensure the continuous control of the installation essential variables. With regard to this complex system problematic, we search the most adapted computer modeling. We show that a multi-agent system approach brings an interesting answer to manage the distribution and parallelism of control decisions and tasks. We present a synthetic study of multi-agent systems and their application fields.The choice of a multi-agent approach proceeds with the design of an agent model. This model gains experiences from other applications. This model is implemented in a computer environment which combines the mechanisms of an object language with Prolog. We propose in this frame a multi-agent modeling of the control system where each function is represented by an agent. The agents are structured in a hierarchical organisation and deal with different abstraction levers of the problem. Following a prototype process, the validation is realized by an implementation and by a coupling to a reactor simulator. The essential contributions of an agent approach turn on the mastery of the system complexity, the openness, the robustness and the potentialities of human-machine cooperation. (author)

  16. Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

    2016-01-15

    A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

  17. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  18. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  19. Bladder Perforation Secondary to Primary Systemic Amyloidosis

    Directory of Open Access Journals (Sweden)

    Christopher J. Dru

    2014-01-01

    Full Text Available Amyloidosis is a disorder of protein folding characterized by extracellular aggregation and deposition of amyloid protein fibrils. Light-chain amyloidosis, also known as primary systemic amyloidosis, is the most common form of the disease. We present a case of an 84-year-old male with a history of systemic primary amyloidosis causing genitourinary, cardiac, and autonomic dysfunction who presented with hematuria and hypotension secondary to bladder perforation. He underwent open repair of a large extraperitoneal bladder defect. He ultimately died as a result of medical complications from his disease.

  20. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  1. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  2. In-situ oxide layer analysis of alloy 182 using electrochemical impedance spectroscopy in high dissolved hydrogen condition in PWR environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho-Sub; Subramanian, Gokul Obulan; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Alloy 82/182 weld metals had been extensively used in joining the components of the PWR primary system. Unfortunately, the cracking caused by PWSCC usually occurs on Alloy 82/182 dissimilar metal welds (DMW). Previous studies indicated that the susceptibility of PWSCC is closely related to the oxide characteristics which are dependent on water chemistry condition, especially dissolved hydrogen (DH). Furthermore, in primary system of pressurized water reactor (PWR), crack initiation resulted from electrochemical instability of oxide film of Ni-base structural materials in various hydrogen concentrations. In this study, in-situ oxide analysis of Alloy 182 using electrochemical impedance spectroscopy (EIS) was performed in high dissolved hydrogen condition. Especially, to understand the effects of tensile loading on the oxide characteristics, we tried to characterize the oxides formed on the tensile loaded specimen using in-situ EIS analysis. The EIS analysis of oxide on Alloy 182 was performed. The increase of oxide film thickness was observed with the increase of exposure time. To analysis the multi-layer structure of oxides, an equivalent model was obtained by fitting EIS data. It is assumed that overall oxide structures were composed of 3 layers approximately.

  3. Primary Angiitis of the Central Nervous System

    Directory of Open Access Journals (Sweden)

    Mojdeh Ghabaee

    2012-03-01

    Full Text Available Primary angiitis of the central nervous system (PACNS is an idiopathic disorder (vasculitis restricted to the central nervous system (CNS. It often presents with focal neurological deficits suggesting stroke or a combination of confusion and headache. We herein report three cases with various combinations of fever, partial seizure, encephalopathy, paresis, headache and ataxia. One of them was initially treated as herpes simplex meningoencephalitis, but further investigations revealed primary angiitis. Primary angiitis of the CNS has protean manifestations and should always be considered in patients suspicious to have CNS infection or stroke, particularly who does not respond to the routine treatments. Clinical data, exclusion of differential diagnoses and typical angiography seem to be enough to justify the diagnosis in the majority of cases.

  4. Primary Angiitis Of The Central Nervous System

    Directory of Open Access Journals (Sweden)

    Sundaram Meenakshi

    2001-01-01

    Full Text Available An unusual case of primary angiitis of central nervous system (PACNS presenting with headache, seizures and focal deficits is presented. Despite multiple lesions noted on brain MRI, definitive diagnosis required a brain biopsy. A high index of clinical suspicious and the utility of brain biopsy for diagnosis are emphasized.

  5. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  6. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  7. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    OpenAIRE

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  8. [Transforming health systems based on primary care].

    Science.gov (United States)

    Durán-Arenas, Luis; Salinas-Escudero, Guillermo; Granados-García, Víctor; Martínez-Valverde, Silvia

    2012-01-01

    Access to health services is a social basic determinant of health in Mexico unlike what happens in developed countries. The demand for health services is focused on primary care, but the design meets only the supply of hospital care services. So it generates a dissonance between the needs and the effective design of health services. In addition, the term affiliation refers to population contributing or in the recruitment process, that has been counted as members of these social security institutions (SS) and Popular Insurance (SP). In the case of Instituto Mexicano del Seguro Social (IMSS) three of four contributors are in contact with health services; while in the SP, this indicator does not exist. Moreover, the access gap between health services is found in the health care packages so that members of the SS and SP do not have same type of coverage. The question is: which model of health care system want the Mexicans? Primary care represents the first choice for increasing the health systems performance, as well as to fulfill their function of social protection: universal access and coverage based on needs, regardless whether it is a public or private health insurance. A central aspect for development of this component is the definition of the first contact with the health system through the creation of a primary health care team, led by a general practitioner as the responsible of a multidisciplinary health team. The process addresses the concepts of primary care nursing, consumption of inputs (mainly medical drugs), maintenance and general services. Adopting a comprehensive strategy that will benefit all Mexicans equally and without discrimination, this primary care system could be financed with a total operating cost of approximately $ 22,809 million by year.

  9. The NIST Primary Radon-222 Measurement System

    OpenAIRE

    Collé, R.; Hutchinson, J. M. R.; Unterweger, M. P.

    1990-01-01

    Within the United States, the national standard for radon measurements is embodied in a primary radon measurement system that has been maintained for over 50 years to accurately measure radon (222Rn) against international and national radium (226Ra) standards. In turn, all of the radon measurements made at the National Institute of Standards and Technology (NIST) and the radon transfer calibration standards and calibration services provided by NIST are directly related to this national radon ...

  10. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  11. Investigation of feedback on neutron kinetics and thermal hydraulics from detailed online fuel behavior modeling during a boron dilution transient in a PWR with the two-way coupled code system DYN3D-TRANSURANUS

    Energy Technology Data Exchange (ETDEWEB)

    Holt, L., E-mail: lars.holt@tuev-sued.de [TÜV SÜD Energietechnik GmbH Baden-Württemberg, Gottlieb-Daimler-Str. 7, 70794 Filderstadt (Germany); Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany); Rohde, U.; Kliem, S.; Baier, S. [Helmholtz-Zentrum Dresden—Rossendorf, Reactor Safety Division, PO Box 510119, D-01314 Dresden (Germany); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstr. 5, D-30457 Hannover (Germany); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Macián-Juan, R. [Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany)

    2016-02-15

    Highlights: • General coupling interface was developed for the fuel performance code TRANSURANUS. • With this new tool simplified fuel behavior models in codes can be replaced. • The reactor dynamics code DYN3D was coupled to TRANSURANUS at assembly level. • The feedback from detailed online fuel behavior modeling is analyzed for reactivity initiated accident (RIA). • The thermal hydraulics can be affected strongly even in fresh fuel assemblies. - Abstract: Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-critical-heat-flux heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m{sup 3} slug of under-borated coolant into a German pressurized water reactor (PWR) core initiated from a sub-critical reactor state (extreme reactivity initiated accident (RIA) conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ 25 J/g, even still around 20 J/g for nodes without film boiling. Furthermore, the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling

  12. Method and means for abruptly terminating the flow of fluid in closed circulating systems of nuclear reactor plants or the like. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Schiele, O.; Florjancic. D.

    1976-08-31

    A nuclear steam supply system is described wherein each of a plurality of centrifugal pumps begins to operate with full cavitation in response to an abrupt drop of system pressure in the event of leakage. This is achieved by influencing the net positive suction head of each pump over the entire range of fluid flow and/or by influencing the net positive suction head upstream of the pumps. The first mode of causing the pumps to operate with full cavitation includes an appropriate selection of the inlet angle and/or inlet diameter of the pump impeller, the provision of auxiliary impellers which are located upstream of the pumps and can circulate the fluid in or counter to the direction of rotation of the respective pump impellers, or the provision of suitably curved guide vanes in the pumps. The second mode include interrupting the admission of undercooled fluid into the system upstream of the pumps.

  13. Study of colloidal particles behaviour in the PWR primary circuit conditions; Etude du comportement des particules colloidales dans les conditions physicochimiques du circuit primaire des reacteurs a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Barale, M

    2006-12-15

    EDF wants to understand, model and limit primary circuit contamination of Pressurized Water Reactors by colloidal particles resulting from corrosion. The electrostatic behaviour of representative oxide particles (cobalt ferrite, nickel ferrite and magnetite) has been studied in primary circuit conditions with the influence of boric acid and lithium hydroxide. The isoelectric point (IEP) and the point of zero charge (PZC) of particles, measured between 5 C and 320 C, exhibit a minimum towards 200 C. The thermodynamic constants of the protonation equilibrium of surface sites were calculated. When boric acid is added, zeta potential and IEP decrease because of borate ions sorption. On the contrary, there is not effect of lithium ions. The modelling of these results under conditions representative of primary circuit shows that these oxides exhibit a negative surface charge, explaining their sorption and adhesion behaviour. (author)

  14. NASAP: a computer code for the evaluation of the Non-proliferation Alternative Systems Assessment Program concepts. Final report in support of Task 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Maul, B.A.

    1979-09-01

    The Non-Proliferation Alternative Systems Assessment Program (NASAP) computer code was developed to calculate the LWR and NASAP choice reactor cost through an arbitrary year T/sub N/. The final cost is arrived at by calculation of cost contributory factors for both LWR and NASAP choice reactors.

  15. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  16. Primary renal osteosarcoma with systemic dissemination

    Directory of Open Access Journals (Sweden)

    Tarun Puri

    2012-01-01

    Full Text Available Primary renal osteosarcoma is an uncommon disease, which, unlike its skeletal counterpart, presents mostly in adults, and is generally diagnosed late due to its non-specific features and intra-abdominal location. Even if the disease is localized at diagnosis, it follows an aggressive course despite radical surgery and adjuvant treatment. We report a case of renal osteosarcoma in a 65-year-old female, who developed regional recurrence, and lung and bone metastases soon after radical nephrectomy for localized disease. Chemotherapy was ineffective in controlling systemic disease.

  17. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  18. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  19. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  20. Development of a parametric containment event tree model of a severe PWR accident

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  1. Horizontal Drop of 21- PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2001-04-26

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

  2. Primary Systemic Amyloidosis with Extensive Gastrointestinal Involvement

    Directory of Open Access Journals (Sweden)

    Vinaya Gaduputi

    2013-12-01

    Full Text Available We report this case of a 42-year-old woman who presented with a debilitating illness manifested by intractable nausea, vomiting, diarrhea and unchecked weight loss. The patient had multisystem involvement that presented as anemia, abnormal liver function tests and progressively deteriorating renal function necessitating dialysis. She was found to be profoundly hypoalbuminemic secondary to malabsorptive and protein-losing enteropathy in tandem with nephrotic range proteinuria. Intolerance to enteral feeding led the patient to be dependent on parenteral nutrition. Serum immunofixation revealed IgG lambda monoclonal protein. The patient underwent endoscopic evaluation with biopsies taken from the gastrointestinal tract that confirmed the diagnosis of primary systemic light-chain amyloidosis. A subsequent bone marrow biopsy revealed normocellular bone marrow with deposition of amyloid. The patient was not considered for autologous stem cell transplantation as the outcomes in patients with multisystem involvement are often poor, with a high mortality risk. Diffuse primary systemic light-chain amyloidosis involving the gastrointestinal tract is a rare entity and is to be considered among differentials in patients presenting with unexplained malabsorptive symptoms.

  3. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  4. [Information system in primary health care].

    Science.gov (United States)

    Stevanović, Ranko; Stanić, Arsen; Varga, Sinisa

    2005-01-01

    The Croatian Ministry of Health started a health care system computerization project aimed at strengthening the collaboration among health care institutions, expert groups and individual health care providers. A tender for informatic system for Primary Health Care (PHC) general practice, pediatrics and gynecology, a vital prerequisite for project realization, has now been closed. Some important reasons for undertaking the project include rationalization of drug utilization, savings through a reduced use of specialists, consultants and hospitalization, then achievement of better cooperation, work distribution, result linking, data quality improvement (by standardization), and ensuring proper information-based decision making. Keeping non-standardized and thus difficult to process data takes too much time of the PHC team time. Since, however, a vast amount of data are collected on only a few indicators, some important information may remain uncovered. Although decisions made by health authorities should rely on evidence and processed information, the authorities spend most of the time working with raw data from which their decisions ultimately derive. The Informatic Technology (IT) in PHC is expected to enable a different approach. PHC teams should be relieved from the tedious task of data gathering and the authorities enabled to work with the information rather than data. The Informatics Communication Technology (ICT) system consists of three parts: hardware (5000 personal computers for work over the Internet), operative system with basic software (editor, etc.), and PHC software for PHC teams. At the national level (National Public Health Informatics System), a software platform will be built for data collection, analysis and distribution. This data collection will be based on the International Classification of Primary Care (ICPC-2) standard to ensure the utilization of medical records and quality assessment. The system permits bi-directional data exchange between

  5. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  6. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  7. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  8. Primary helium heater for propellant pressurization systems

    Science.gov (United States)

    Reichmuth, D. M.; Nguyen, T. V.; Pieper, J. L.

    1991-01-01

    The primary helium heater is a unique design that provides direct heating of pressurant gas for large pressure fed propulsion systems. It has been conceptually designed to supply a heated (800-1000 R) pressurization gas to both a liquid oxygen and an RP-1 propellant tank. This pressurization gas is generated within the heater by mixing super critical helium (40-300 R and 3000-1600 psi) with an appropriate amount of combustion products from a 4:1 throttling stoichiometric LO2/LH2 combustor. This simple, low cost and reliable mixer utilizes the large quantity of helium to provide stoichiometric combustor cooling, extend the throttling limits and enhance the combustion stability margin. Preliminary combustion, thermal, and CFD analyses confirm that this low-pressure-drop direct helium heater can provide the constant-temperature pressurant suitable for tank pressurization of both fuel and oxidizer tanks of large pressure fed vehicles.

  9. Advances in Primary Central Nervous System Lymphoma.

    Science.gov (United States)

    Patrick, Lauren B; Mohile, Nimish A

    2015-12-01

    Primary central nervous system lymphoma (PCNSL) is a rare form of non-Hodgkin lymphoma that is limited to the CNS. Although novel imaging techniques aid in discriminating lymphoma from other brain tumors, definitive diagnosis requires brain biopsy, vitreoretinal biopsy, or cerebrospinal fluid analysis. Survival rates in clinical studies have improved over the past 20 years due to the addition of high-dose methotrexate-based chemotherapy regimens to whole-brain radiotherapy. Long-term survival, however, is complicated by clinically devastating delayed neurotoxicity. Newer regimens are attempting to reduce or eliminate radiotherapy from first-line treatment with chemotherapy dose intensification. Significant advances have also been made in the fields of pathobiology and treatment, with more targeted treatments on the horizon. The rarity of the disease makes conducting of prospective clinical trials challenging, requiring collaborative efforts between institutions. This review highlights recent advances in the biology, detection, and treatment of PCNSL in immunocompetent patients.

  10. Dynamic modeling of primary and secondary systems of IRIS reactor for transient analysis using SIMULINK

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Lima, Fernando Roberto de Andrade, E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2011-07-01

    The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)

  11. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

  12. Procedure qualification of CNP650 PWR primary coolant pipeline by manual welding%CNP650型压水堆主管道手工焊接工艺评定

    Institute of Scientific and Technical Information of China (English)

    刘先文

    2012-01-01

    The primary coolant pipe of CNP650 pressurized water reactor is the enterclose of coolant of reactor core,which is the pressure pipe of large diameter and thickness connected with RPV(reactor pressure vessel) and SG(steam generator) and RCP(reactor coolant pump).The welding construction of primary coolant pipe is the pivotal path of the installation of main equipment of nuclear island and the key and difficult point of the construction of nuclear power plant.The data sheet and welding experience of the WPQ is very important for ensuring the success of the first welding construction.The process control of the manual WPQ of CNP650 nuclear power plant of QinShan Nuclear Power Phase II Expansion Project included the simulation condition of the site and the management of welding process and physical and chemical testing and welding deformation, in order to get the deposited metal fitting for the requirements of the NDE and physical and chemical properties of the technical specification.The process control is the prerequisite of the welding construction of primary coolant pipe.%CNP650型压水堆的主管道作为反应堆压力容器堆芯冷却剂的通道,是连接反应堆压力容器、主泵和蒸汽发生器的大型厚壁承压管道.主管道焊接施工是核岛主设备安装的关键路径,是核电建设的重点与难点.焊接工艺评定所提供的数据与焊接经验,对确保主管道焊接施工一次成功,起着非常重要的作用.泰山核电二期扩建工程CNP650型核电站主管道手工焊接工艺评定从模拟现场焊接施工的条件、焊接过程管理、理化试验、焊接变形等方面进行控制,以获得符合技术规范对熔敷金属无损检测、理化性能的要求.焊接工艺评定过程控制为主管道焊接施工提供先决条件.

  13. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  14. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  15. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  16. Degraded core analysis for the PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  17. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  18. Coupled Analysis on Steady Flow in Primary and Secondary Sides of PWR Steam Generator%压水堆蒸汽发生器一、二次侧稳态流场耦合分析

    Institute of Scientific and Technical Information of China (English)

    丛腾龙; 田文喜; 秋穗正; 苏光辉; 谢永诚; 姚彦贵

    2014-01-01

    The steam generator (SG) suffers from the challenge of tube rupture caused by the flow induced vibration (FIV ) during operation .T he 3D flow characteristics in SG are essential for the analysis of FIV .The secondary side flow field was simulated based on the porous media model ,using the coupled heat transfer from primary to secondary side . T he 3D velocity , temperature , pressure and quality distributions in secondary side ,the one-dimensional temperature and heat transfer coefficient (HTC) distributions of primary and secondary sides as well as the U-tube temperature distribution were obtained .The thermal-hydraulic characteristic distributions in the shell side and the flow vapor quality distribution in separators are significantly uneven due to the non-uniformly distributed heat source released from primary to secondary side .The flow quality distribution in the separators varies from 0.05 to 0.62 ,w hich can be used to the design for separator load .The velocity distribution in the U-bend region was calculated , w hich provides input conditions for the evaluation of FIV damage of tubes .%蒸汽发生器(S G )在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。

  19. Development of Educational Management System in Small Primary School

    Science.gov (United States)

    Alsammarry, Yupayao; Sirisuthi, Chaiyuth; Duangcharthom, Surat

    2016-01-01

    The purposes of the research were: (1) to study the factors of Educational Management System in Small Primary School; (2) to investigate current situations problems and guidelines of developing educational management in small primary school; (3) to develop Educational Management System in Small Primary School; and (4) to examine the results of…

  20. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  1. Diversity of primary care systems analysed.

    NARCIS (Netherlands)

    Kringos, D.; Boerma, W.; Bourgueil, Y.; Cartier, T.; Dedeu, T.; Hasvold, T.; Hutchinson, A.; Lember, M.; Oleszczyk, M.; Pavlick, D.R.

    2015-01-01

    This chapter analyses differences between countries and explains why countries differ regarding the structure and process of primary care. The components of primary care strength that are used in the analyses are health policy-making, workforce development and in the care process itself (see Fig.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  3. Deep primary production in coastal pelagic systems

    DEFF Research Database (Denmark)

    Lyngsgaard, Maren Moltke; Richardson, Katherine; Markager, Stiig

    2014-01-01

    produced. The primary production (PP) occurring below the surface layer, i.e. in the pycnocline-bottom layer (PBL), is shown to contribute significantly to total PP. Oxygen concentrations in the PBL are shown to correlate significantly with the deep primary production (DPP) as well as with salinity...... that eutrophication effects may include changes in the structure of planktonic food webs and element cycling in the water column, both brought about through an altered vertical distribution of PP....

  4. RELAP5 Analyses of ROSA/LSTF Experiments on AM Measures during PWR Vessel Bottom Small-Break LOCAs with Gas Inflow

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2014-01-01

    Full Text Available RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW injection into the secondary-side of both steam generators (SGs as the AM measures were taken 10 min after a safety injection signal. The primary depressurization rate of 55 K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses.

  5. Teachers' Performance Motivation System in Thai Primary Schools

    Science.gov (United States)

    Pasathang, Sarojn; Tesaputa, Kowat; Sataphonwong, Pattananusron

    2016-01-01

    This research aims to: 1) study the present conditions and desirable condition of the motivation systems as well as how to find methods for motivating the performance of teachers in primary schools, 2) develop a motivation system for the performance of teachers in primary schools, 3) study the effects of using the motivation system for compliance…

  6. Pediatric Primary Care as a Component of Systems of Care

    Science.gov (United States)

    Brown, Jonathan D.

    2010-01-01

    Systems of care should be defined in a manner that includes primary care. The current definition of systems of care shares several attributes with the definition of primary care: both are defined as community-based services that are accessible, accountable, comprehensive, coordinated, culturally competent, and family focused. However, systems of…

  7. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  8. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  9. The advanced main control console for next japanese PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  10. Measuring the strength of primary care systems in Europe.

    NARCIS (Netherlands)

    Kringos, D.S.; Boerma, W.G.W.

    2009-01-01

    Background: The investment in primary care (PC) reforms to improve the overall performance of health care systems has been substantial in Europe. There is however a lack of up to date comparable information to evaluate the development and strength of PC systems. This EU-funded Primary Health Care A

  11. A sustainable primary care system: lessons from the Netherlands.

    NARCIS (Netherlands)

    Faber, M.J.; Burgers, J.S.; Westert, G.P.

    2012-01-01

    The Dutch primary care system has drawn international attention, because of its high performance at low cost. Primary care practices are easily accessible during office hours and collaborate in a unique out-of-hours system. After the reforms in 2006, there are no copayments for patients receiving ca

  12. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  13. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  14. VERIFIKASI KECELAKAAN HILANGNYA ALIRAN AIR UMPAN PADA REAKTOR DAYA PWR MAJU

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Reactor Technology and Nuclear Safety as a Technical Support Organization (TSO in terms of reactor safety verification, the verification activities have been carried out for the AP1000 that begins with failure of secondary coolant accident verification. The activity started with the technical safety features modeling such as passive core cooling system consisting of a Passive Residual Heat Removal system (PRHR, Core Makeup Tank (CMT, and In-containment Refueling Water Storage Tank (IRWST. The failure of secondary coolant accident selected is the loss of main feedwater flow to one of the steam generator simulated using the calculation program RELAP5/SCDAP/Mod3.4. The objective of analysis is to obtain sequences of changes in the thermalhydraulic parameters in the reactor due to the selected event. Analysis results obtained are validated and compared with the analysis results using the calculation program LOFTRAN in the AP1000 safety design document. The verification results show that the loss of feed-water supply has no impact on core damage, the reactor coolant system, as well as secondary systems. The ability of heat exchanger PRHR has been verified to dissipate decay heat of the core after reactor trip. Validation with the AP1000 safety design document shows compliance on most thermal hydraulic parameters. In general, the advanced PWR model equipped with passive core cooling system that has been developed remains safe in the event of loss of secondary coolant flow accident. Keywords: Verification, loss of feed water flow, AP1000

  15. School Management Information Systems in Primary Schools

    Science.gov (United States)

    Demir, Kamile

    2006-01-01

    Developments in information technologies have been impacting upon educational organizations. Principals have been using management information systems to improve the efficiency of administrative services. The aim of this research is to explore principals' perceptions about management information systems and how school management information…

  16. Fuzzy control applied to nuclear power plant pressurizer system

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Mauro V.; Almeida, Jose C.S., E-mail: mvitor@ien.gov.b, E-mail: jcsa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In a pressurized water reactor (PWR) nuclear power plants (NPPs) the pressure control in the primary loop is very important for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes, during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study has two main goals: first is to develop a pressurizer model based on artificial neural networks (ANNs); second is to develop a fuzzy controller for the PWR pressurizer pressure, and compare its performance with the P controller. Data from a simulator PWR plant was used to test the ANN and the controllers as well. The reference simulator is a Westinghouse 3-loop PWR plant with a total thermal output of 2785 MWth. The simulation results show that the pressurizer ANN model response are in reasonable agreement with the simulated power plant, and the fuzzy controller built in this study has better performance compared to the P controller. (author)

  17. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  18. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  19. Application of Total Quality Management System in Thai Primary Schools

    Science.gov (United States)

    Prueangphitchayathon, Setthiya; Tesaputa, Kowat; Somprach, Kanokorn

    2015-01-01

    The present study seeks to develop a total quality management (TQM) system that can be applied to primary schools. The approach focuses on customer orientation, total involvement of all constituencies and continuous improvement. TQM principles were studied and synthesized according to case studies of the best practices in 3 primary schools (small,…

  20. Application of Total Quality Management System in Thai Primary Schools

    Science.gov (United States)

    Prueangphitchayathon, Setthiya; Tesaputa, Kowat; Somprach, Kanokorn

    2015-01-01

    The present study seeks to develop a total quality management (TQM) system that can be applied to primary schools. The approach focuses on customer orientation, total involvement of all constituencies and continuous improvement. TQM principles were studied and synthesized according to case studies of the best practices in 3 primary schools (small,…

  1. Primary energy-transformations in biological systems

    Energy Technology Data Exchange (ETDEWEB)

    Lehninger, A.L.

    1980-10-01

    In this paper I shall review the main outlines of current research on the molecular aspects of the primary energy-coupling mechanisms in cells, those carried out by energy-transducing membranes. They include the capture of solar energy by the chloroplast membranes of green plants, used to generate carbohydrates and molecular oxygen from carbon dioxide and water, and the counterpart of photosynthesis, the process of respiration in heterotrophic organisms, in which reduced organic products generated by photosynthesis are oxidized at the expense of dioxygen to form carbon dioxide and water. Although the cycling of dioxygen, carbon dioxide, and organic matter between the plant and animal worlds is well known, it is not generally appreciated that the magnitude of biological energy flux in these cycles is huge compared to the total energy flux in man-made devices. A major consequence is that the concentration of carbon dioxide in the atmosphere has been increasing at a significant rate, at a time when there is also a decrease, at least in some parts of the world, in the counterbalancing utilization of CO/sub 2/ by green plants, due to deforestation. The greenhouse effect of increased atmospheric CO/sub 2/ may not only change the earth's climate, but also may influence the rate of photosynthesis. It is also not generally appreciated that energy flow in the biosphere leads to production of enormous amounts of organic matter potentially useful in furnishing man's energy requirements.

  2. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Directory of Open Access Journals (Sweden)

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  3. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  4. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Energy Technology Data Exchange (ETDEWEB)

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

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

  5. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  6. PWR fuel in Japan; The changes and trend for hereafter

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki

    1992-07-01

    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.).

  7. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  8. Zinc injection in German PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B. [Framatome ANP GmbH, Erlangen (Germany); Juergensen, M. [Kernkraftwerk Obrigheim GmbH (Germany); Wolter, D. [RWE Power AG, Kraftwerk Biblis (Germany)

    2002-07-01

    Zinc injection for further reduction of radiation fields was introduced at Unit B of Biblis Nuclear Power Station in September 1996 and at Obrigheim Nuclear Power Station in February 1998. Zinc injection is still being implemented today at these plants. This paper gives an overview of the experience acquired with the method, including the annual refueling outages in the year 2001. The main topic addressed by the paper is the evolution of dose rates at the primary system and work-related doses since introduction of the method. Reductions in high dose rate areas have meanwhile achieved values of 40 to 50%. Annual collective doses per man-hour spent in the controlled access area of the plant as well as personal doses for specific activities are also decreasing. (authors)

  9. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment

    Science.gov (United States)

    Zhang, Litao; Wang, Jianqiu

    2014-03-01

    Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics.

  10. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  11. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Science.gov (United States)

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo

    2016-04-01

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

  12. Analyses of PWR boron dilution consequences with the Arrotta code

    Energy Technology Data Exchange (ETDEWEB)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  13. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  14. ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2016-01-01

    Full Text Available Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility. The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

  15. Application of a PID controller based on fuzzy logic to reduce variations in the control parameters in PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Wagner Eustaquio de; Lira, Carlos Alberto Brayner de Oliveira; Brito, Thiago Souza Pereira de; Afonso, Antonio Claudio Marques, E-mail: wagner@unicap.br, E-mail: cabol@ufpe.br, E-mail: afonsofisica@gmail.com, E-mail: thiago.brito86@yahoo.com.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Cruz Filho, Antonio Jose da; Marques, Jose Antonio, E-mail: antonio.jscf@gmail.com, E-mail: jamarkss@uol.com.br [Universidade Catolica de Pernambuco (CCT/PUC-PE), Recife, PE (Brazil). Centro de Ciencias e Tecnologia; Teixeira, Marcello Goulart, E-mail: marcellogt@dcc.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Matematica. Dept. de Matematica

    2013-07-01

    Nuclear reactors are in nature nonlinear systems and their parameters vary with time as a function of power level. These characteristics must be considered if large power variations occur in power plant operational regimes, such as in load-following conditions. A PWR reactor has a component called pressurizer, whose function is to supply the necessary high pressure for its operation and to contain pressure variations in the primary cooling system. The use of control systems capable of reducing fast variations of the operation variables and to maintain the stability of this system is of fundamental importance. The best-known controllers used in industrial control processes are proportional-integral-derivative (PID) controllers due to their simple structure and robust performance in a wide range of operating conditions. However, designing a fuzzy controller is seen to be a much less difficult task. Once a Fuzzy Logic controller is designed for a particular set of parameters of the nonlinear element, it yields satisfactory performance for a range of these parameters. The objective of this work is to develop fuzzy proportional-integral-derivative (fuzzy-PID) control strategies to control the level of water in the reactor. In the study of the pressurizer, several computer codes are used to simulate its dynamic behavior. At the fuzzy-PID control strategy, the fuzzy logic controller is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region. Thus the fuzzy logic controller tunes the gain of PID controller to adapt the model with changes in the water level of reactor. The simulation results showed a favorable performance with the use to fuzzy-PID controllers. (author)

  16. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  17. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  18. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  19. PWR fuel in Japan; Progress and future trends

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))

    1994-06-01

    Twenty years ago, in the early years of the Japanese civil nuclear power programme, the fuel used was imported from Westinghouse in the USA. However, it was always intended that there would be a move towards fuel fabrication in Japan and by the end of 1993 around 10,000 Mitsubishi PWR fuel assemblies had been supplied to 21 PWRs in Japan. The highest burnup achieved so far is 46 GWd/t. Design changes to reduce abnormalities have been made, reliability is improving all the time and further improvements in burnup are being developed. This progress in PWR cores and fuel including MOX fuel in Japan is charted and future research and development is outlined. (UK).

  20. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  1. Control of corrosion product transport in PWR secondary cycles

    Energy Technology Data Exchange (ETDEWEB)

    Sawochka, S.G.; Pearl, W.L. [NWT Corp., San Josa, CA (United States); Passell, T.O.; Welty, C.S. [Electric Power Research Institute, Palo Alto, CA (United States)

    1992-12-31

    Transport of corrosion products to PWR steam generators by the feedwater leads to sludge buildup on the tubesheets and fouling of tube-to-tube support crevices. In these regions, chemical impurities concentrate and accelerate tubing corrosion. Deposit buildup on the tubes also can lead to power generation limitations and necessitate chemical cleaning. Extensive corrosion product transport data for PWR secondary cycles has been developed employing integrating sampling techniques which facilitate identification of major corrosion product sources and assessments of the effectiveness of various control options. Plant data currently are available for assessing the impact of factors such as pH, pH control additive, materials of construction, blowdown, condensate treatment, and high temperature drains and feedwater filtration.

  2. Binary mass ratios: system mass not primary mass

    CERN Document Server

    Goodwin, Simon P

    2012-01-01

    Binary properties are usually expressed (for good observational reasons) as a function of primary mass. It has been found that the distribution of companion masses -- the mass ratio distribution -- is different for different primary masses. We argue that system mass is the more fundamental physical parameter to use. We show that if system masses are drawn from a log-normal mass function, then the different observed mass ratio distributions as a function of primary mass, from M-dwarfs to A-stars, are all consistent with a universal, flat, system mass ratio distribution. We also show that the brown dwarf mass ratio distribution is not drawn from the same flat distribution, suggesting that the process which decides upon mass ratios is very different in brown dwarfs and stars.

  3. Decommissioning of the BR3 PWR

    Energy Technology Data Exchange (ETDEWEB)

    Massaut, V.; Klein, M

    1998-07-01

    The objectives, programme and main achievements of SCK-CEN's decommissioning programme in 1997 are summarised. Particular emphasis is on the BR3 decommissioning project. In 1997, auxiliary equipment and loops were dismantled; concrete antimissile slabs were decontaminated; the radiology of the primary loop was modelled; the quality assurance procedure for dismantling loops and equipment were implemented; a method for the dismantling of the reactor pressure vessel was selected; and contaminated thermal insulation of the primary loop containing asbestos was removed.

  4. The 747 primary flight control systems reliability and maintenance study

    Science.gov (United States)

    1979-01-01

    The major operational characteristics of the 747 Primary Flight Control Systems (PFCS) are described. Results of reliability analysis for separate control functions are presented. The analysis makes use of a NASA computer program which calculates reliability of redundant systems. Costs for maintaining the 747 PFCS in airline service are assessed. The reliabilities and cost will provide a baseline for use in trade studies of future flight control system design.

  5. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  6. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  7. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    James, L.A.

    1997-10-01

    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  8. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen

    2005-07-01

    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

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

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

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

    Science.gov (United States)

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

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  11. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  12. PWR core stablity aganst xenon-induced spatial power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))

    1982-06-01

    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  13. Actinides transmutation - a comparison of results for PWR benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  14. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  15. The Cuban National Healthcare System: Characterization of primary healthcare services.

    Directory of Open Access Journals (Sweden)

    Keli Regina DAL PRÁ

    2015-10-01

    Full Text Available This article presents a report on the experience of healthcare professionals in Florianópolis, who took the course La Atención Primaria de Salud y la Medicina Familiar en Cuba [Primary Healthcare and Family Medicine in Cuba], in 2014. The purpose of the study is to characterize the healthcare units and services provided by the Cuban National Healthcare System (SNS and to reflect on this experience/immersion, particularly on Cuba’s Primary Healthcare Service. The results found that in comparison with Brazil’s Single Healthcare System (SUS Cuba’s SNS Family Healthcare (SF service is the central organizing element of the Primary Healthcare Service. The number of SF teams per inhabitant is different than in Brazil; the programs given priority in the APS are similar to those in Brazil and the intersectorial nature and scope of the services prove to be effective in the resolution of healthcare problems.

  16. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  17. A system look at electromechanical actuation for primary flight control

    NARCIS (Netherlands)

    Lomonova, E.A.

    1997-01-01

    An overview is presented of the emergence of the ALL Electric flight control system (FCS) or power-by-wire (PBW) concept. The concept of fly-by-power refers to the actuator using electrical rather than hydraulic power. The development of the primary flight control Electromechanical Actuators (EMAs)

  18. The integration of public health in European primary care systems.

    NARCIS (Netherlands)

    Kringos, D.S.; Bolibar, Y.; Bourgueil, T.; Cartier, T.; Dedeum, T.; Hasvold, A.; Hutchinson, M.; Lember, M.; Oleszczyk, D.; Rotar Pavlick, I.; Svab, P.; Tedeschi, A.; Wilson, S.; Wilm, A.; Windak, A.; Boerma, W.

    2010-01-01

    Background: A strong primary care (PC) system provides accessible, comprehensive care in an ambulatory setting on a continuous basis and by coordinated care processes. These features give PC the opportunity to play a key role in providing public health (PH) services to their practice population. Th

  19. The integration of public health in European primary care systems.

    NARCIS (Netherlands)

    Kringos, D.S.; Bolibar, Y.; Bourgueil, T.; Cartier, T.; Dedeum, T.; Hasvold, A.; Hutchinson, M.; Lember, M.; Oleszczyk, D.; Rotar Pavlick, I.; Svab, P.; Tedeschi, A.; Wilson, S.; Wilm, A.; Windak, A.; Boerma, W.

    2010-01-01

    Background: A strong primary care (PC) system provides accessible, comprehensive care in an ambulatory setting on a continuous basis and by coordinated care processes. These features give PC the opportunity to play a key role in providing public health (PH) services to their practice population. Th

  20. A system look at electromechanical actuation for primary flight control

    NARCIS (Netherlands)

    Lomonova, E.A.

    1997-01-01

    An overview is presented of the emergence of the ALL Electric flight control system (FCS) or power-by-wire (PBW) concept. The concept of fly-by-power refers to the actuator using electrical rather than hydraulic power. The development of the primary flight control Electromechanical Actuators (EMAs)

  1. Enlarged tongue due to primary systemic amyloidosis:clinicopathologic observation

    Institute of Scientific and Technical Information of China (English)

    潘卫红; 李娜萍; 梁国芬

    2004-01-01

    @@Primary amyloidosis (AL) is characterized by deposition of abnormal extra cellular protein in the form of fibrils in many organs, especially the heart, kidneys, gastrointestinal tract, and peripheral nervous system.1 Involvement of the tongue is not uncommon in primary AL. In 22 % to 26 % of patients suffering from AL, amyloid deposition in the tongue can result in an enlarged tongue.2 Lingual changes arising from localization or from all over, AL has been reported frequently as initial signs of the disease. Macroscopically, it is difficult to distinguish AL of the tongue from other lesions. In view of the variety of protein species involved and the wide spectrum of possible clinical presentations,3 the diagnosis of lingual AL is frequently overlooked, because immunohistochemical studies of such cases have not been undertaken. Here we describe two male patients with primary systemic AL who developed enlarged tongues. In addition to the manifestation of lingual AL, in which oral signs were the primary indicators of the disease, we describe the immunohistochemical findings of the tongue to discuss diagnostic criteria for lingual amyloid in primary AL.

  2. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Energy Technology Data Exchange (ETDEWEB)

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  3. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)

    2013-07-01

    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  4. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  5. Targetable genetic features of primary testicular and primary central nervous system lymphomas.

    Science.gov (United States)

    Chapuy, Bjoern; Roemer, Margaretha G M; Stewart, Chip; Tan, Yuxiang; Abo, Ryan P; Zhang, Liye; Dunford, Andrew J; Meredith, David M; Thorner, Aaron R; Jordanova, Ekaterina S; Liu, Gang; Feuerhake, Friedrich; Ducar, Matthew D; Illerhaus, Gerald; Gusenleitner, Daniel; Linden, Erica A; Sun, Heather H; Homer, Heather; Aono, Miyuki; Pinkus, Geraldine S; Ligon, Azra H; Ligon, Keith L; Ferry, Judith A; Freeman, Gordon J; van Hummelen, Paul; Golub, Todd R; Getz, Gad; Rodig, Scott J; de Jong, Daphne; Monti, Stefano; Shipp, Margaret A

    2016-02-18

    Primary central nervous system lymphomas (PCNSLs) and primary testicular lymphomas (PTLs) are extranodal large B-cell lymphomas (LBCLs) with inferior responses to current empiric treatment regimens. To identify targetable genetic features of PCNSL and PTL, we characterized their recurrent somatic mutations, chromosomal rearrangements, copy number alterations (CNAs), and associated driver genes, and compared these comprehensive genetic signatures to those of diffuse LBCL and primary mediastinal large B-cell lymphoma (PMBL). These studies identify unique combinations of genetic alterations in discrete LBCL subtypes and subtype-selective bases for targeted therapy. PCNSLs and PTLs frequently exhibit genomic instability, and near-uniform, often biallelic, CDKN2A loss with rare TP53 mutations. PCNSLs and PTLs also use multiple genetic mechanisms to target key genes and pathways and exhibit near-uniform oncogenic Toll-like receptor signaling as a result of MYD88 mutation and/or NFKBIZ amplification, frequent concurrent B-cell receptor pathway activation, and deregulation of BCL6. Of great interest, PCNSLs and PTLs also have frequent 9p24.1/PD-L1/PD-L2 CNAs and additional translocations of these loci, structural bases of immune evasion that are shared with PMBL. © 2016 by The American Society of Hematology.

  6. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  7. Primary central nervous system lymphoma in an immunocompetent patient

    OpenAIRE

    Málaga-Zenteno, José; Médico Asistente, Servicio de Hematología, Hospital Nacional Carlos Alberto Seguín Escobedo, EsSalud, Arequipa, Perú.; Mamani-Quispe, Jersson Alonso; Estudiante de Medicina Humana, Centro de Investigación y Estudios Médicos (CIEM), Universidad Católica Santa María, Arequipa, Perú. Sociedad Científica Médico Estudiantil Peruana (SOCIMEP).; Fuentes Fuentes, Mariela; Médico Asistente, Servicio de Hematología, Hospital Nacional Carlos Alberto Seguín Escobedo, EsSalud, Arequipa, Perú.; Suclla-Velásquez, José Alonso; Estudiante de Medicina Humana, Centro de Investigación y Estudios Médicos (CIEM), Universidad Católica Santa María, Arequipa, Perú. Sociedad Científica Médico Estudiantil Peruana (SOCIMEP).; Meza Aragón, Julio; Médico Asistente, Servicio de Neurocirugía, Hospital Nacional Carlos Alberto Seguín Escobedo, EsSalud, Arequipa, Perú.

    2012-01-01

    Primary central nervous system lymphoma (PCNSL) constitutes 2% of extranodal lymphomas and 0,3%-1,5% of all intracranial neoplasms in immunocompetent patients, being more frequent after the sixth decade of life. We report a case of a 76 year-old man with no antecedents who started his disease with march instability, difficulty to move left side of his body with brachial predominance, holocraneal headache and dizziness. He arrived at emergency with Glasgow 14 and right eyelid ptosis. He had le...

  8. The Possibility of Building Nuclear Power Plant Free from Severe Accident Risk PWR NPP with advanced all passive safety cooling systems (AAP SCS)%发展无严重事故风险核电站的曙光具有完全非能动安全冷却系统的压水堆核电站

    Institute of Scientific and Technical Information of China (English)

    肖宏才

    2013-01-01

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP,actuated by natural force has been put forward in the article.Here the natural force mainly means the fore,which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another,including occurrence of accident situation.Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident,so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink.There is no need to rely on automatic control system,any active equipment and human actions in all working process of the AAP SCS,which can reduce the probability of severe accident to zero,so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety.Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology.So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk,and for modernization of existing second generation nuclear power plant.%本文提出了用自然力直接触发启动压水堆核电站一整套完全非能动的停堆安全冷却系统.这里的自然力主要是指一回路运行工况转换时由于其压力分布变化所形成的压差力.在这一系统中,当进行停堆或发生某种一回路事故工况时,相应的安全冷却系统便自然地投入运行,立即缓解事故后果,将事故时一回路释放的能量及堆芯余热非能动地排入最终热阱.在全过程中不依靠自动控制系统、能动设备及任何人为因素的介入,即可确保对堆芯余热无限期的安全冷却能力,完全避免压水堆核电站发生向环境泄漏放射性物

  9. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  10. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-29

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

  11. RELAP5 MODEL OF THE DIVERTOR PRIMARY HEAT TRANSFER SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    Popov, Emilian L [ORNL; Yoder Jr, Graydon L [ORNL; Kim, Seokho H [ORNL

    2010-08-01

    This report describes the RELAP5 model that has been developed for the divertor primary heat transfer system (PHTS). The model is intended to be used to examine the transient performance of the divertor PHTS and evaluate control schemes necessary to maintain parameters within acceptable limits during transients. Some preliminary results are presented to show the maturity of the model and examine general divertor PHTS transient behavior. The model can be used as a starting point for developing transient modeling capability, including control system modeling, safety evaluations, etc., and is not intended to represent the final divertor PHTS design. Preliminary calculations using the models indicate that during normal pulsed operation, present pressurizer controls may not be sufficient to keep system pressures within their desired range. Additional divertor PHTS and control system design efforts may be required to ensure system pressure fluctuation during normal operation remains within specified limits.

  12. Preoperational test report, primary ventilation condenser cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-10-29

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  13. Behavior modification in primary care: the pressure system model.

    Science.gov (United States)

    Katz, D L

    2001-01-01

    The leading causes of death in the United States are predominantly attributable to modifiable behaviors. Patients with behavioral risk factors for premature death and disability, including dietary practices; sexual practices; level of physical activity; motor vehi cle use patterns; and tobacco, alcohol, and illicit sub stance use, are seen far more consistently by primary care providers than by mental health specialists. Yet models of behavior modification are reported, debated, and revised almost exclusively in the psychology literature. While the Stages of Change Model, or Transtheo retical Model, has won application in a broadening array of clinical settings, its application in the primary care setting is apparently quite limited despite evidence of its utility [Prochaska J, Velicer W. Am J Health Promot 1997;12:38-48]. The lack of a rigorous behavioral model developed for application in the primary care setting is an impediment to the accomplishment of public health goals specified in the Healthy People objectives and in the reports of the U.S. Preventive Services Task Force. The Pressure System Model reported here synthesizes elements of established behavior modification theories for specific application under the constraints of the primary care setting. Use of the model in both clinical and research settings, with outcome evaluation, is encouraged as part of an effort to advance public health.

  14. Radiative heat transfer modelling in a PWR severe accident sequence

    Energy Technology Data Exchange (ETDEWEB)

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    Full text of publication follows: The present study is devoted to the estimation of the radiative heat transfers during a severe accident sequence in a Pressurized Water Reactor. In such a situation, the residual nuclear power released by the fuel rods can not be evacuated and heats up the core. As a result, the cylindrical rods and the structures initially composing the core undergo a degradation process: swelling, breaking or melting of the rods and structures and eventual collapse to form a heap of fragments called a debris bed. As the solid matrix loses its original shape, the core geometry continuously evolves from standing, regularly-spaced cylinders to a non-homogeneous system including deformed remaining rods and structures and debris particles. To predict this type of sequence, the ICARE/CATHARE software [1] is developed by IRSN. Since the temperatures can reach values greater than 3000 K, it was of major interest to provide the code with an accurate radiative transfer model usable whatever the geometry of the system. Considering the size of a reactor core compared to the mean penetration length of radiation, the core can be seen as an optically thick medium. This observation led us to use the diffusion approximation to treat the radiation propagation. In this approach, the radiative flux is calculated in a way similar to thermal conduction: q{sub r} = [K{sub e}].{nabla}T where [K{sub e}] is the equivalent conductivity tensor of the system accounting for thermal and radiative transfer. An homogenization technique is applied to estimate the equivalent conductivity. Given the temperature level, the radiative contribution to the equivalent conductivity tensor quickly becomes dominant. This model was described earlier in [2] in which it was shown that an equivalent conductivity can be continuously calculated in the system when the geometry evolves from standing regular cylinder rods to swollen or broken ones, surrounded or not by a film of liquid materials, to

  15. Experience on Primary System Decommissioning in Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

    Paloma Molleda; Leandro Sanchez; David Rodriguez [ENSA, Cantabria (Spain)

    2015-10-15

    Primary System Decommissioning belongs to DCP(Decommissioning and Closure Plan) works and its scope includes: Steam Generator, Pressurizer, Refrigerant Circuit Pump and Primary Circuit Piping. All these dismantling activities were carried out on site, including preliminary steps before their removal (SAS installations, pre decontaminations, cutting and segmentations, segregations, etc.) and delivery to media/low activity nuclear waste disposal site. There are many cutting techniques available in market (most of them proved with positive results) as well as there are many different approaches about how to manage radioactive wastes in decommissioning projects (containers or great components disposal, containers burial, re fusion, etc.). Both issues are linked and, before starting a new project, it might be positive and quite useful to compare and study previous dismantling experiences, especially the lesson learned chapter. Primary System cut with diamond saw has been a challenge target, not only due to the methodology innovation (since until nowadays, the common use of this technology was performed in cutting concrete walls) because it has a huge range of positive aspects that, in our opinion, are attractive (apart from its mentioned versatility, in terms of cutting on site and every type of material)

  16. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  17. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  18. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  19. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  20. EVALUATION OF THE TEMPORARY TENT COVER TRUSS SYSTEM AP PRIMARY VENT SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    HAQ MA

    2009-12-31

    The purpose of this calculation is to evaluate a temporary ten cover truss system. This system will be used to provide weather protection to the workers during replacement of the filter for the Primary Ventilation System in AP Tank Farm. The truss system has been fabricated utilizing tubes and couplers, which are normally used for scaffoldings.

  1. Strategies to reduce PWR inspection time

    Energy Technology Data Exchange (ETDEWEB)

    Guerra, J.; Gonzalez, E. [TECNATOM SA, Madrid (Spain)

    2001-07-01

    During last few years, a constant reduction in inspection time was clearly demanded by most nuclear plant owners. This requirement has to be accomplished without any impact in inspection quality that, in general, has also to be improved. All this in a market with increasing competition that forces price reductions. Under these new demands from our customers, Tecnatom reoriented its development efforts to improve his products and services to meet this challenges. Two of our main inspection activities that have clear impact in outage duration are Steam Generator and Vessel inspections. This paper describes the improvements made in these two activities as an example of the reorientation of our development efforts with a focus on the technical improvements made on the software and robotic tools applied as in the data acquisition and analysis systems. In the Steam Generator inspections, new robots with dual guide tubes are commonly used. New eddy current instruments and software were developed to keep up with the data rates produced by the faster acquisition system. Use of automatic analysis software is also helping to improve speed while reducing cost and improving overall job quality. Production rates are close to double from the previous inspection system. (author)

  2. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Energy Technology Data Exchange (ETDEWEB)

    Gourguechon, B.

    1993-12-31

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

  3. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

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

    Energy Technology Data Exchange (ETDEWEB)

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  5. CAREM: an innovative-integrated PWR

    Energy Technology Data Exchange (ETDEWEB)

    Mazzi, R. [INVAP Nuclear Projects Div., Rio Negro (Argentina)], E-mail: mazzi@invap.com.ar

    2009-07-01

    Presented on March 1984 in an international conference for the first time, 'CAREM Concept' focused on engineering solutions from early stages of the design that minimize requirements to safety and safeguards systems making the product simpler, highly reliable and cost effective. The overall idea was widely adopted by worldwide designers, originated a new category of small a medium size nuclear power plants frequently know as 'integrated reactor' and/or 'Advanced-passive safety-reactor'. This paper describes the main design features, progress and prospects of the CAREM project as well as proliferation resistant conditions applicable to the design. (author)

  6. Masquerade Syndrome of Multicentre Primary Central Nervous System Lymphoma

    Directory of Open Access Journals (Sweden)

    Silvana Guerriero

    2011-01-01

    Full Text Available Purpose. In Italy we say that the most unlucky things can happen to physicians when they get sick, despite the attention of colleagues. To confirm this rumor, we report the sad story of a surgeon with bilateral vitreitis and glaucoma unresponsive to traditional therapies. Methods/Design. Case report. Results. After one year of steroidal and immunosuppressive therapy, a vitrectomy, and a trabeculectomy for unresponsive bilateral vitreitis and glaucoma, MRI showed a multicentre primary central nervous system lymphoma, which was the underlying cause of the masquerade syndrome. Conclusions. All ophthalmologists and clinicians must be aware of masquerade syndromes, in order to avoid delays in diagnosis.

  7. Primary anaplastic large T cell lymphoma of central nervous system

    Directory of Open Access Journals (Sweden)

    ZHANG Yan

    2013-01-01

    Full Text Available Background Primary anaplastic large T cell lymphoma (ALCL of central nervous system (CNS can occur in people of all ages, and is usually unrelated with immunodeficiency. It is often misdiagnosed as meningitis, especially tuberculous meningitis, on clinical practice and imaging examination. In pathological diagnosis, the morphological changes of primary ALCL of CNS are similar to the systemic ALCL and the anaplastic lymphoma kinase-1 (ALK-1 can be positive or negative. Being misdiagnosed as meningitis, hormone therapy with glucocorticoid before biopsy is always used, and massive necrosis and a lot of histocyte proliferation and phagocytosis can be found under histological findings. Therefore, when the material is not enough, primary ALCL of CNS is often misdiagnosed as cerebral infarction or malignant histocytosis and so on. This paper reports a case of primary ALCL of CNS and makes a review of relevant literature, so as to summarize the clinical manifestations and elevate the recognition of clinicians and pathologists on this disease. Methods and Results A 12-year-old boy was admitted because of fever, worsening headache, numbness and weakness of right limbs. MRI showed local gyri swelling and abnormal enhancement of pia mater in the right parietal lobe, expanding to the right temporal lobe, and pia mater enhancement in the left parietal lobe. The right temporo-parietal lobe lesion biopsy revealed irregularly shaped tumor cells of large size, rich and eosinophilic cytoplasm and horseshoe-shaped or kidney-shaped nuclei. Immunohistochemical examination showed tumor cells positive for CD3, CD45RO, CD30, ALK-1 and epithelial membrane antigen (EMA, and negative for CD20 and CD79a. Conclusion Primary ALCL of CNS is an extremely rare tumor which is usually misdiagnosed as meningitis according to clinical and imaging examinations. Therefore, for those patients who are considered as meningitis but with poor treatment effect and replase of illness, brain

  8. Analysis of SBO ATWS for Maanshan PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

    2015-11-15

    Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

  9. Timing analysis of PWR fuel pin failures

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Straka, M. (Halliburton NUS, Idaho Falls, ID (United States))

    1992-09-01

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

  10. Development of an allergy management support system in primary care

    Science.gov (United States)

    Flokstra - de Blok, Bertine MJ; van der Molen, Thys; Christoffers, Wianda A; Kocks, Janwillem WH; Oei, Richard L; Oude Elberink, Joanne NG; Roerdink, Emmy M; Schuttelaar, Marie Louise; van der Velde, Jantina L; Brakel, Thecla M; Dubois, Anthony EJ

    2017-01-01

    Background Management of allergic patients in the population is becoming more difficult because of increases in both complexity and prevalence. Although general practitioners (GPs) are expected to play an important role in the care of allergic patients, they often feel ill-equipped for this task. Therefore, the aim of this study was to develop an allergy management support system (AMSS) for primary care. Methods Through literature review, interviewing and testing in secondary and primary care patients, an allergy history questionnaire was constructed by allergists, dermatologists, GPs and researchers based on primary care and specialists’ allergy guidelines and their clinical knowledge. Patterns of AMSS questionnaire responses and specific immunoglobulin E (sIgE)-test outcomes were used to identify diagnostic categories and develop corresponding management recommendations. Validity of the AMSS was investigated by comparing specialist (gold standard) and AMSS diagnostic categories. Results The two-page patient-completed AMSS questionnaire consists of 12 (mainly) multiple choice questions on symptoms, triggers, severity and medication. Based on the AMSS questionnaires and sIgE-test outcome of 118 patients, approximately 150 diagnostic categories of allergic rhinitis, asthma, atopic dermatitis, anaphylaxis, food allergy, hymenoptera allergy and other allergies were identified, and the corresponding management recommendations were formulated. The agreement between the allergy specialists’ assessments and the AMSS was 69.2% (CI 67.2–71.2). Conclusion Using a systematic approach, it was possible to develop an AMSS that allows for the formulation of diagnostic and management recommendations for GPs managing allergic patients. The AMSS thus holds promise for the improvement of the quality of primary care for this increasing group of patients. PMID:28352197

  11. Evaluation of low-level radioactive waste activity in the primary system and auxiliary systems of PWR reactors for purposes of nuclear decommissioning; Avaliacao da atividade de residuos de baixa no sistema primario e sistemas auxiliares de reatores PWR com propositos de desmantelamento nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Melo, Andre T.S.; Maiorino, Jose R., E-mail: andre.melo@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.b [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2013-07-01

    This work will focus on the study of the deposition of Co-60 on the surface of stainless steel (SS-304), based on empirical studies and experimental data. The temporal evolution of the concentration of Co-60 (μCi / cm) will be reported, as well as qualitative discussion about the mechanisms of this deposition.

  12. An evaluation of four telemedicine systems for primary care.

    Science.gov (United States)

    Dunn, E V; Conrath, D W; Bloor, W G; Tranquada, B

    1977-01-01

    In an evaluation of the efficacy of four two-way telecommunication systems for use in primary care, more than 1,000 patients seeking care at a community health center received an additional remote examination by use of either color television, black and white television, still-frame black and white television, or hands-free telephone. The diagnosis, clinical tests and X rays requested, and proposed patient management were compared to the actual care received by the patients at the health center. There were no significant differences between any of the modes in relation to diagnostic accuracy, time for the diagnostic interview, tests requested, or referral rates. Furthermore, patient attitudes did not vary significantly. Thus the relatively inexpensive telephone proved to be as efficient and effective a means for delivery of remote physician care as did any of the visual communication systems. PMID:873812

  13. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  14. French experience in transient data collection and fatigue monitoring of PWR`s nuclear steam supply system; Experience francaise sur la comptabilisation des transitoires et la surveillance en fatigue des chaudieres REP

    Energy Technology Data Exchange (ETDEWEB)

    Sabaton, M.; Morilhat, P.; Savoldelli, D.; Genette, P.

    1995-10-01

    Electricite de France (EDF), the french national electricity company, is operating 54 standardized pressurizer water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYsteme de Surveillance en FAtigue de la Chaudiere) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After the first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF`s nuclear 900 MW is planned after one more year of feedback experience. (authors). 12 refs., 3 figs.

  15. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  16. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  17. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  18. Identification and evaluation of PWR in-vessel severe accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  19. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    Energy Technology Data Exchange (ETDEWEB)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  20. Integrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Yang, Ming; Zhang, Zhijian [Harbin Engineering University, Harbin (China)

    2014-08-15

    The development of a new risk monitor system is introduced in this paper, which can be applied not only to severe accident prevention in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The summary of the fundamental method is summarized on how to configure the Plant Defense in-Depth (Did) Risk Monitor by object-oriented software system based on functional modeling approach. Following the authors??preceding preliminary study for AP1000, the way of realizing the proposed method of configuring the plant Did risk monitor was investigated for a safety-enhanced Japanese PWR design to meet with the tight anti-severe accident requirements set by national regulation in Japan after Fukushima Daiichi accident. The result of this example practice of the presented preliminary study for Japanese PWR was for the level 4 of the Did in case of beyond design basis accident, that is, loss of all AC power + RCP seal LOCA, against the former case of AP1000 for level 3 Did in case of large LOCA.

  1. Neuropsychological profile of patients with primary systemic hypertension.

    Science.gov (United States)

    Ostrosky-Solis, F; Mendoza, V U; Ardila, A

    2001-01-01

    Arterial hypertension represents a risk factor for cerebrovascular disease. It has been hypothesized that chronic hypertension may eventually result in small subcortical infarcts associated with some cognitive impairments. One hundred fourteen patients with primary systemic hypertension (PSH) and 114 matched subjects were selected. PSH patients were further divided in four groups depending upon the hypertension severity. In addition to the medical and laboratory exams, a neuropsychological evaluation was administered. The NEUROPSI neuropsychological test battery was used. An association between level of hypertension and cognitive impairment was observed. Most significant differences were observed in the following domains: Reading, executive functioning, constructional, and memory-recall. No differences were observed in orientation, memory-recognition, and language. Some neuropsychological functions appeared impaired even in the PSH group with the least risk factors. Cognitive evaluation may be important in cases of PSH not only to determine early subtle cognitive changes, but also for follow-up purposes, and to assess the efficacy of different therapeutic procedures.

  2. Primary Histiocytic Sarcoma of the Central Nervous System

    Science.gov (United States)

    So, Hoonsub; Kim, Sun A; Yoon, Dok Hyun; Khang, Shin Kwang; Hwang, Jihye; Suh, Chong Hyun; Suh, Cheolwon

    2015-01-01

    Histiocytic sarcoma is a type of lymphoma that rarely involves the central nervous system (CNS). Its rarity can easily lead to a misdiagnosis. We describe a patient with primary CNS histocytic sarcoma involving the cerebral hemisphere and spinal cord, who had been initially misdiagnosed as demyelinating disease. Two biopsies were necessary before a correct diagnosis was made. A histologic examination showed bizarre shaped histiocytes with larger nuclei and nuclear atypia. The cells were positive for CD68, CD163, and S-100 protein. As a resection was not feasible due to multifocality, he was treated with highdose methotrexate, but showed no response. As a result, he was switched to high dose cytarabine; but again, showed no response. The patient died 2 months from the start of chemotherapy and 8 months from the onset of symptoms. Since few patients with this condition have been described and histopathology is difficult to diagnose, suspicion of the disease is essential. PMID:25345462

  3. 49 CFR 214.529 - In-service failure of primary braking system.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false In-service failure of primary braking system. 214... Maintenance Machines and Hi-Rail Vehicles § 214.529 In-service failure of primary braking system. (a) In the event of a total in-service failure of its primary braking system, an on-track roadway...

  4. PRIMARY CENTRAL NERVOUS SYSTEM LYMPHOMA: CLINICOPATHOLOGICAL AND IMMUNOHISTOCHEMICAL PROFILE

    Directory of Open Access Journals (Sweden)

    Kanwardeep Singh

    2016-03-01

    Full Text Available BACKGROUND Primary central nervous system lymphoma (PCNSL is a rare form of extranodal non-Hodgkin lymphoma (NHL confined to the brain, spinal cord and/or eye, occurring in immunocompetent individuals. Histologically, they are diffuse large B-cell lymphomas. Over the last few decades there has been a gradual increase in their incidence. AIM To study the clinical, histopathological and immunohistochemical profile of primary central nervous system lymphoma. SETTING AND DESIGN Retrospective audit of seven cases of PCNSL diagnosed over a period of five years in a tertiary referral hospital of North India. MATERIAL AND METHODS The clinical, radiological and laboratory findings were retrieved from the hospital records. Histopathology slides were reviewed, studied in detail and a panel of immunohistochemical markers comprising of CD3, CD5, CD20, CD10, BCL6, BCL2, MUM1, CD30, EBV (LMP1, Ki-67 and p53 was done on all cases. RESULTS The male to female ratio was 3:4 with a median age of 60 years. The most common form of presentation was neurological deficits and altered sensorium. Imaging showed contrast enhancing, single or multiple, deep seated lesions within the cerebral hemispheres. Histologically, all were high-grade diffuse large B-cell lymphomas showing typical angiocentricity and a median Ki-67 proliferative index of 80%. Based on immunohistochemistry (Hans classifier three cases had germinal centre B-cell (GCB and four had non-germinal centre B-cell (non-GCB phenotype. p53 was expressed in all cases with strong expression in four of them. Four patients died before treatment could be initiated, one received palliative chemo-radiotherapy and two did not follow up after diagnosis. CONCLUSIONS Primary CNS lymphomas are high-grade diffuse large B-cell lymphomas which show high Ki-67 proliferative indices and frequent overexpression of p53. Irrespective of histological subtype, GCB or non-GCB, outcome is uniformly poor. Early and prompt diagnosis is

  5. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  6. COMPARISON EDUCATION SYSTEMS OF DENMARK AND TURKEY AT PRIMARY LEVEL

    Directory of Open Access Journals (Sweden)

    Deniz YÜCEER

    2012-03-01

    Full Text Available Research is made according to the following problems: “How is education system for primary school in Denmark?” and “In what ways show Turkey?” In this search it is used the horizontal and descriptive approach as a method. For this purpose it is compared similarities and differences by examining the relevant literature in identifying approach; it is also discussed all dimensions in education system, the general structure of education system, funding, courses in primary school, duration of courses, learning environment, assessment etc. In this study the primary and secondary sources are examined. It is used documents, programs (Folkeskole that obtained from Denmark Ministry of Education and researcher’s observations as the primary sources; it is utilized comparative education literature as the secondary sources. Findings: Education is free in both countries. Education is compulsory between the ages of 6-7 and 16-17 years in Denmark; it is also compulsory between the ages 6-14 in Turkey. It is educate active citizens the general goals of the education system in both countries. It is made EVA (Danish Evaluation Institute which is an independent structure of Ministry Education the evaluation about education system. In Turkey inspectors who work within the Ministry of Education are controlled teachers and administrators at the education and training level. Whether the education is received in a publicly provided school, in a private school or at home is a matter of individual choice, as long as accepted standards are met. Also there are centers of municipality which spend leisure time for children. It is also not found centers of this kind in Turkey. The Folkeskole consists of one year of pre-school class, nine years of primary and lower secondary education and one-year 10th class and 10th class is optional. Primary schools and pre-school education is called as "Folkeskole”. Content is distributed within three subject areas as social

  7. Recommendations from primary care providers for integrating mental health in a primary care system in rural Nepal.

    Science.gov (United States)

    Acharya, Bibhav; Tenpa, Jasmine; Thapa, Poshan; Gauchan, Bikash; Citrin, David; Ekstrand, Maria

    2016-09-19

    Globally, access to mental healthcare is often lacking in rural, low-resource settings. Mental healthcare services integration in primary care settings is a key intervention to address this gap. A common strategy includes embedding mental healthcare workers on-site, and receiving consultation from an off-site psychiatrist. Primary care provider perspectives are important for successful program implementation. We conducted three focus groups with all 24 primary care providers at a district-level hospital in rural Nepal. We asked participants about their concerns and recommendations for an integrated mental healthcare delivery program. They were also asked about current practices in seeking referral for patients with mental illness. We collected data using structured notes and analyzed the data by template coding to develop themes around concerns and recommendations for an integrated program. Participants noted that the current referral system included sending patients to the nearest psychiatrist who is 14 h away. Participants did not think this was effective, and stated that integrating mental health into the existing primary care setting would be ideal. Their major concerns about a proposed program included workplace hierarchies between mental healthcare workers and other clinicians, impact of staff turnover on patients, reliability of an off-site consultant psychiatrist, and ability of on-site primary care providers to screen patients and follow recommendations from an off-site psychiatrist. Their suggestions included training a few existing primary care providers as dedicated mental healthcare workers, recruiting both senior and junior mental healthcare workers to ensure retention, recruiting academic psychiatrists for reliability, and training all primary care providers to appropriately screen for mental illness and follow recommendations from the psychiatrist. Primary care providers in rural Nepal reported the failure of the current system of referral, which

  8. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    Energy Technology Data Exchange (ETDEWEB)

    Mayers, J.B.; Soth, L.G.

    1978-04-01

    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.

  9. Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

    Full Text Available The Pennsylvania State University (PSU under the sponsorship of the US Nuclear Regulatory Commission (NRC has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC benchmark based on the Nuclear Power Engineering Corporation (NUPEC pressurized water reactor (PWR subchannel and bundle tests (PSBTs. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA of OECD and the Japan Nuclear Energy Safety Organization (JNES, Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB, under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.

  10. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  11. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  12. Mutational analysis of primary central nervous system lymphoma.

    Science.gov (United States)

    Bruno, Aurélie; Boisselier, Blandine; Labreche, Karim; Marie, Yannick; Polivka, Marc; Jouvet, Anne; Adam, Clovis; Figarella-Branger, Dominique; Miquel, Catherine; Eimer, Sandrine; Houillier, Caroline; Soussain, Carole; Mokhtari, Karima; Daveau, Romain; Hoang-Xuan, Khê

    2014-07-15

    Little is known about the genomic basis of primary central nervous system lymphoma (PCNSL) tumorigenesis. To investigate the mutational profile of PCNSL, we analyzed nine paired tumor and germline DNA samples from PCNSL patients by high throughput exome sequencing. Eight genes of interest have been further investigated by focused resequencing in 28 additional PCNSL tumors to better estimate their incidence. Our study identified recurrent somatic mutations in 37 genes, some involved in key signaling pathways such as NFKB, B cell differentiation and cell cycle control. Focused resequencing in the larger cohort revealed high mutation rates for genes already described as mutated in PCNSL such as MYD88 (38%), CD79B (30%), PIM1 (22%) and TBL1XR1 (19%) and for genes not previously reported to be involved in PCNSL tumorigenesis such as ETV6 (16%), IRF4 (14%), IRF2BP2 (11%) and EBF1 (11%). Of note, only 3 somatically acquired SNVs were annotated in the COSMIC database. Our results demonstrate a high genetic heterogeneity of PCNSL and mutational pattern similarities with extracerebral diffuse large B cell lymphomas, particularly of the activated B-cell (ABC) subtype, suggesting shared underlying biological mechanisms. The present study provides new insights into the mutational profile of PCNSL and potential targets for therapeutic strategies.

  13. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  14. Cognitive Multiple Access Network with Outage Margin in the Primary System

    CERN Document Server

    Maham, Behrouz; Zhou, Xiangyun; Hjørungnes, Are

    2011-01-01

    This paper investigates the problem of spectrally efficient operation of a multiuser uplink cognitive radio system in the presence of a single primary link. The secondary system applies opportunistic interference cancelation (OIC) and decode the primary signal when such an opportunity is created. We derive the achievable rate in the secondary system when OIC is used. This scheme has a practical significance, since it enables rate adaptation without requiring any action from the primary system. The \\emph{exact} expressions for outage probability of the primary user are derived, when the primary system is exposed to interference from secondary users. Moreover, approximated formulas and tight lower and upper bounds for the ergodic sum-rate capacity of the secondary network are found. Next, the power allocation is investigated in the secondary system for maximizing the sum-rate under an outage constraint at the primary system. We formulate the power optimization problem in various scenarios depending on the avail...

  15. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Directory of Open Access Journals (Sweden)

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  16. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  17. Design and Implementation of SCADA System Based Power Distribution for Primary Substation (Control System

    Directory of Open Access Journals (Sweden)

    Khin Thu Zar Win

    2014-10-01

    Full Text Available SCADA stands for Supervisory Control and Data Acquisition. SCADA system is more porpular than other control system in the modern industrial processes. This research describes the automated switch control for SCADA based electrical distribution system of primary substation by using PLC. The objective of this research is to transform the manual control system to automated switch control system in Myanmar. There are four main portions in SCADA based electrical distribution system. They are automated control system, interfacing units, monitoring system and networking system. The automated control system is emphasised in this research. This system can be accomplished by using PLC ladder diagram. This automated distribution system is analyzed to develop a secure, reliabe and convenient management tool which can use remote terminal units (RTUs. The simulations based approach automated system are demonstrated in this research. According to the simulation results, the proposed automated control system using PLC are met with the desired control environment with high performance stage. This system is efficient and reliable for conventional electrical distribution system in Myanmar by using SCADA based technology.

  18. Primary Sjögren's syndrome as a systemic disease

    DEFF Research Database (Denmark)

    Malladi, Arundathi S; Sack, Kenneth E; Shiboski, Stephen C;

    2012-01-01

    To study the prevalence of extraglandular manifestations in primary Sjögren's syndrome (SS) among participants enrolled in the Sjögren's International Collaborative Clinical Alliance (SICCA) Registry....

  19. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  20. Performance of monosphere new gel type ion exchange resins for condensate polisher at PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Nakanishi, S.; Nakamura, M.; Asou, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Izumi, T.; Deguchi, T.; Ino, T.; Hagiwara, M.

    1998-12-31

    There are two kinds of ion exchange resins of gel type and porous one which are used as condensate polisher in LWR nuclear power plants. In order to estimate the performance of these resins on the condensate polisher at the secondary cycle of Japanese PWR plants, a column test was performed setting the column test device in Ohi power station unit 1 of the Kansai Electric Power Co., Inc. and the variations of the resin properties and the samples at the end of column were analyzed. The column test showed that the cross-linking degree of the new gel resins used was lower than those of porous ones. The new resins captured larger amounts of Matrix-Diffused Crud than the conventional cation resins before regeneration but not after that. Whereas the surface adsorbed crud was less captured by the new resins than conventional anion resins. However, there were little differences among these resins in respects of rinsing characteristics, sphericity, water quality, break through capacity, etc. At the condensate polisher in the secondary system it was confirmed that new gel resins had almost the same performance as one of the conventional ones and could be applied to the actual plant. (M.N.)

  1. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  2. Micromorphology of Restorative System-Dentin Interface in Primary Teeth Using Different Adhesive Systems and Burs

    Directory of Open Access Journals (Sweden)

    Tereza Cristina Favieri de MELO-SILVA

    2007-03-01

    Full Text Available Objective: This study in vitro evaluated the micromorphology of the resin-dentin interface in primary teeth, using different rotatory instruments and adhesive systems. Method: Twenty primary molars were selected and randomly divided into two groups (n=10. In group C, the occlusal surfaces of teeth were cut with a carbide bur at high-speed until the area of dentin exposure. In group D, the same procedure was conducted, but the dentin was cut with a diamond bur. The surface of each tooth was divided into two halves; one half of the occlusal surface received application of two-step total-etch adhesive system (Single-Bond – 3M, and the other half received application of one-step self-etching adhesive (One Up Bond F - Tokuyama. All teeth were restored with composite (Z-250 - 3M. The samples were thermocycled, embedded in acrylic resin, sectioned for achievement of the resin-dentin interface and the teeth were sputter- coated with gold and observed under SEM. Results: the two adhesive systems showed hybrid layer formation; the two-step adhesive system demonstrated better interface sealing than the self-etching; the dentin cut with carbide burs was not statistically different with regard to the adhesive system; and diamond bur with self-etching adhesive system showed the worst interface sealing with the highest gap values. Conclusion: the diamond bur presented negative influence only in the quality of the interface between restorative system and primary dentin when it was used the one step self-etching adhesive system.

  3. Impact analysis of different operation strategies for battery energy storage systems (BESS) providing primary control reserve

    OpenAIRE

    Fleer, Johannes; Stenzel, Peter; Linssen, Jochen

    2015-01-01

    In this work, a techno-economic analysis of stationary battery systems providing primary control for grid stabilisation is conducted. The effects of battery design and operation strategies adapted for primary control supply are investigated with regard to costs and parameters relevant for battery aging. Primary control is required to balance the feed-in and use of electricity to/from the grid, thereby ensuring safe and stable grid operation. In Germany, primary control is traded on a separate...

  4. Misunderstanding and Understanding of Primary Water Stress Corrosion Cracking of Structural Components in the Primary System of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kim, Sung Soo; Kim, Dae Whan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    All the structural components in the primary system of pressurized water reactors that are in contact with primary water are made of austenitic Ni-Cr-Fe alloys which are known to be corrosion resistant. Nevertheless, these Ni-Cr-Fe alloys such as Alloy 600, weld 182/82, austenitic stainless steels suffer from intergranular stress corrosion cracking (IGSCC) after their 10 year operation in reactors although the environment to which they have been exposed is almost pure water of pH 6.9 to 7.2, which is called primary water stress corrosion cracking (PWSCC). Given that the underlying mechanism of PWSCC remains unidentified so far, there are many misunderstandings related to PWSCC of the structural components, which may lead to unreasonable mitigation measures. The aim of this work is to highlight understanding and misunderstanding of PWSCC related to austenitic Ni-Cr-Fe alloys.

  5. General digitalized system on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Akagi, Katsumi; Kadohara, Hozumi; Taniguchi, Manabu [Mitsubishi Electric Corp., Tokyo (Japan)

    2000-08-01

    Hitherto, instrumentation control system in a PWR nuclear power plant has stepwisely adopted digital technology such as application of digital instrumentation control device to ordinary use (primary/secondary system control device, and so on), application of CRT display system to monitoring function, and so forth, to realize load reduction of an operator due to expansion of operation automation range, upgrading of reliability and maintenance due to self-diagnosis function, reduction of mass in cables due to multiple transfer, and upgrading of visual recognition due to information integration. In next term PWR plant instrumentation control system, under consideration of application practice of conventional digital technology, application of general digitalisation system to adopt digitalisation of overall instrumentation control system containing safety protection system, and central instrumentation system (new type of instrumentation system) and to intend to further upgrade economics, maintenance, operability/monitoring under security of reliability/safety is planned. And, together with embodiment of construction program of the next-term plant, verification at the general digitalisation proto-system aiming at establishment of basic technology on the system is carried out. Then, here was described on abstract of the general digitalisation system and characteristics of a digital type safety protection apparatus to be adopted in the next-term plant. (G.K.)

  6. Primary Nocturnal Enuresis: A Structural and Strategic Family Systems Approach.

    Science.gov (United States)

    Fletcher, Teresa B.

    2000-01-01

    Exploration of the literature regarding primary nocturnal enuresis suggests there are various causes including genetic, biological, physiological, and psychological explanations. Treatments typically consist of medication and behavioral intervention. However, it was believed that this enuretic case was caused by psychological trauma. A series of…

  7. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  8. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  9. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  10. Continuous firefly algorithm applied to PWR core pattern enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Poursalehi, N., E-mail: npsalehi@yahoo.com [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-05-15

    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

  11. French nuclear plants PWR vessel integrity assessment and life management

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  12. Instrumentation Needs for Integral Primary System Reactors (IPSRs) - Task 1 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Gary D. Storrick; Bojan Petrovic; Luca Oriani; Lawrence E. Conway; Diego Conti

    2005-09-30

    This report presents the results of the Westinghouse work performed under Task 1 of this Financial Assistance Award and satisfies a Level 2 Milestone for the project. While most of the signals required for control of IPSRs are typical of other PWRs, the integral configuration poses some new challenges in the design or deployment of the sensors/instrumentation and, in some cases, requires completely new approaches. In response to this consideration, the overall objective of Task 1 was to establish the instrumentation needs for integral reactors, provide a review of the existing solutions where available, and, identify research and development needs to be addressed to enable successful deployment of IPSRs. The starting point for this study was to review and synthesize general characteristics of integral reactors, and then to focus on a specific design. Due to the maturity of its design and availability of design information to Westinghouse, IRIS (International Reactor Innovative and Secure) was selected for this purpose. The report is organized as follows. Section 1 is an overview. Section 2 provides background information on several representative IPSRs, including IRIS. A review of the IRIS safety features and its protection and control systems is used as a mechanism to ensure that all critical safety-related instrumentation needs are addressed in this study. Additionally, IRIS systems are compared against those of current advanced PWRs. The scope of this study is then limited to those systems where differences exist, since, otherwise, the current technology already provides an acceptable solution. Section 3 provides a detailed discussion on instrumentation needs for the representative IPSR (IRIS) with detailed qualitative and quantitative requirements summarized in the exhaustive table included as Appendix A. Section 3 also provides an evaluation of the current technology and the instrumentation used for measurement of required parameters in current PWRs. Section 4

  13. Cognitive Multiple-Antenna Network with Outage and Rate Margins at the Primary System

    DEFF Research Database (Denmark)

    Maham, Behrouz; Popovski, Petar

    2015-01-01

    In the common model for spectrum sharing, cognitive users can access the spectrum as long as the target performance in the legitimate primary system is not violated. In this paper, we consider a downlink primary multiple-inputsingle- output (MISO) system which operates under a controlled interfer......In the common model for spectrum sharing, cognitive users can access the spectrum as long as the target performance in the legitimate primary system is not violated. In this paper, we consider a downlink primary multiple-inputsingle- output (MISO) system which operates under a controlled...... interference from the downlink MISO cognitive radio, also called secondary system. We derive exact expressions for outage probability of the primary user under Rayleigh fading, when the primary system is exposed to interference from a secondary base station. We treat three different operating modes...... of the adaptive-rate transmit-beamforming primary system and the adaptive-rate transmit antenna-selection primary system. The analytical results are confirmed by simulations, in which we analyze the impact of different parameters, such as the number of antennas and the amount of the interference on the system...

  14. Cognitive multiple-antenna network in outage-restricted primary system

    DEFF Research Database (Denmark)

    Maham, Behrouz; Popovski, Petar

    2013-01-01

    In the commons model for the spectrum sharing, cognitive users can access the spectrum as long as the target performance in the legitimate primary system is not violated. In this paper, we consider a downlink primary multiple-input-single-output (MISO) system which operates under a controlled...... interference from the downlink MISO cognitive radio, also called secondary system. We derive exact expressions for outage probability of the primary user under Rayleigh fading, when the primary system is exposed to interference from a secondary base station. Moreover, in high-SNR scenario, a closed...... there is an outage constraint at the primary system, and a simple solution is proposed. Finally, the analytical results are confirmed by simulations, in which we analyze the impact of different parameters, such as the number of antennas and the amount of the interference on the system performance; these could...

  15. SYSTEM OF DISTANCE LEARNING ADMINISTRATION IN CONTINUING EDUCATION FOR PRIMARY SCHOOL TEACHERS

    OpenAIRE

    MUKOVIZ, Oleksii P.

    2015-01-01

    The article describes the peculiarities of the organization of primary school teachers continuing education by means of web technologies, presents the website of the system of primary school teachers continuing education (http://sno.udpu.org.ua), and analyzes its content and structure. The website of the system of primary school teachers continuing education is created with the help of four instrumental platforms WordPress, Moodle, PhpBB and “cloud” technologies from Microsoft (SkyDrive, Padl...

  16. Primary and Secondary Congestion Management in Restructured Power Systems

    Directory of Open Access Journals (Sweden)

    Mehdi Hajian

    2015-03-01

    Full Text Available In this paper a new strategy for congestion management is presented. Generally, in congestion management programs all considered contingencies are deterministic or stochastic. Therefore, congestion management cost is increased. In the proposed algorithm, congestion management program is divided to two parts, Primary and Secondary Congestion Management. Also, contingencies are divided in two groups. In the first group, the contingencies that should be managed by preventive actions are considered and the others are taken into account in the second group. The first group of contingencies is in primary congestion management. The secondary congestion management program starts from the nearest time to market run time. So, the secondary congestion management uses corrective actions for managing secondary group contingencies. Also, the secondary congestion management is capable of covering most uncertainties, especially load variations. The proposed method reduces congestion management cost.

  17. Capturing the complexity of European primary care systems in a European monitoring instrument.

    NARCIS (Netherlands)

    Kringos, D.; Boerma, W.

    2009-01-01

    Aim: The investment in PC reforms to improve the overall performance of health care systems has been substantial in Europe. There is however a lack of up to date comparable information to evaluate the development of primary care (PC) systems. This EU-funded PHAMEU (Primary Health Care Activity Monit

  18. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

  19. Primary central nervous system angiosarcoma: two case reports

    Directory of Open Access Journals (Sweden)

    Hackney James R

    2012-08-01

    Full Text Available Abstract Introduction Primary angiosarcoma of the brain is extremely rare; only 15 cases have been reported in adults over the last 25 years. Case presentations We describe two cases of primary angiosarcoma of the brain that are well characterized by imaging, histopathology, and immunohistochemistry. Case 1: our first patient was a 35-year-old woman who developed exophthalmos. Subtotal resection of a left extra-axial retro-orbital mass was performed. Case 2: our second patient was a 47-year-old man who presented to our facility with acute visual loss, word-finding difficulty and subtle memory loss. A heterogeneously-enhancing left sphenoid wing mass was removed. We also review the literature aiming at developing a rational approach to diagnosis and treatment, given the rarity of this entity. Conclusions Gross total resection is the standard of care for primary angiosarcoma of the brain. Adjuvant radiation and chemotherapy are playing increasingly recognized roles in the therapy of these rare tumors.

  20. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  1. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    Energy Technology Data Exchange (ETDEWEB)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  2. On-line PWR RHR pump performance testing following motor and impeller replacement

    Energy Technology Data Exchange (ETDEWEB)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  3. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K [ORNL; Sung, Yixing [Westinghouse Electric Company, Cranberry Township; Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township; Xu, Yiban [Westinghouse Electric Company, Cranberry Township; Cao, Liping [Westinghouse Electric Company, Cranberry Township

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  4. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  5. Description of the primary flight display and flight guidance system logic in the NASA B-737 transport systems research vehicle

    Science.gov (United States)

    Knox, Charles E.

    1990-01-01

    A primary flight display format was integrated with the flight guidance and control system logic in support of various flight tests conducted with the NASA Transport Systems Research Vehicle B-737-100 airplane. The functional operation of the flight guidance mode control panel and the corresponding primary flight display formats are presented.

  6. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  7. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  8. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  9. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables. (ACR)

  10. Primary energy savings using heat storage for biomass heating systems

    Directory of Open Access Journals (Sweden)

    Mitrović Dejan M.

    2012-01-01

    Full Text Available District heating is an efficient way to provide heat to residential, tertiary and industrial users. The heat storage unit is an insulated water tank that absorbs surplus heat from the boiler. The stored heat in the heat storage unit makes it possible to heat even when the boiler is not working, thus increasing the heating efficiency. In order to save primary energy (fuel, the boiler operates on nominal load every time it is in operation (for the purpose of this research. The aim of this paper is to analyze the water temperature variation in the heat storage, depending on the heat load and the heat storage volume. Heat load is calculated for three reference days, with average daily temperatures from -5 to 5°C. The primary energy savings are also calculated for those days in the case of using heat storage in district heating.[Projekat Ministarstva nauke Republike Srbije, br. TR 33051: The concept of sustainable energy supply of settlements with energy efficient buildings

  11. The W. M. Keck Telescope segmented primary mirror active control system

    Energy Technology Data Exchange (ETDEWEB)

    Jared, R.C.; Arthur, A.A.; Andreae, S.; Biocca, A.; Cohen, R.W.; Fuertes, J.M.; Franck, J.; Gabor, G.; Llacer, J.; Mast, T.; Meng, J.; Merrick, T.; Minor, R.; Nelson, J.; Orayani, M.; Salz, P.; Schaefer, B.; Witebsky, C.

    1989-07-01

    The ten meter diameter primary mirror of the W. M. Keck Telescope is a mosaic of thirty-six hexagonal mirrors. An active control system stabilizes the primary mirror. The active control system uses 168 measurements of the relative positions of adjacent mirror segments and 3 measurements of the primary mirror position in the telescope structure to control the 108 degrees of freedom needed to stabilize the figure and position of the primary mirror. The components of the active control system are relative position sensors, electronics, computers, actuators that position the mirrors, and software. The software algorithms control the primary mirror, perform star image stacking, emulate the segments, store and fit calibration data, and locate hardware defects. We give an overview of the active control system, its functional requirements and test measurements. 12 refs.

  12. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  13. Primary reserve studies for high wind power penetrated systems

    DEFF Research Database (Denmark)

    Das, Kaushik; Altin, Müfit; Hansen, Anca Daniela;

    2015-01-01

    With high penetration of non-synchronous wind generations replacing conventional generators, the inertia of power system will reduce. A large disturbance in such a power system can cause faster frequency change in this power system and might invoke emergency defence strategies like underfrequency....... This paper further explores the capabilities of wind turbines to provide support during underfrequency to prevent load shedding. Maximum wind penetration possible without causing load shedding following a large disturbance is also investigated....

  14. Atypical Imaging Findings in Primary Central Nervous System Lymphoma

    Directory of Open Access Journals (Sweden)

    Zahra Afravi

    2010-05-01

    Full Text Available Background/Objective: The incidence of primary CNS lymphomas (PCNSL is increasing. Timely diagnosis of PCNSL can lead to proper therapeutic management. There are some atypical imaging findings that may easily be misdiagnosed as other pathologic processes such as infectious and demyelinative diseases. As a result, histopathologic diagnosis is necessary for all suspected lesions."nPatients and Methods: In this research we studied 120 cases of PCNSL over the past 16 years. Some of them had atypical imaging findings, suggesting many differential diagnoses. Having said that, stereotactic biopsy was performed for all cases and the diagnosis was proved."nResults: We selected some interesting cases with atypical imaging findings of PCNSL, which were unlikely to be diagnosed without histopathologic evaluation. "nConclusion: PCNSL must be kept in mind as a differential diagnosis for other brain lesions. Histopathologic diagnosis is necessary for prompt management.

  15. Can casemix-systems be applied in Danish primary care?

    DEFF Research Database (Denmark)

    Halling, Anders; Kristensen, Troels

    Background: New technology in terms of IT systems, better data infrastructure and improved registrations of health data provide new opportunities for health care systems to improve the care experience of individual patients, improve public health and reduce healthcare costs. Application of "Big...

  16. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  17. Intention and Usage of Computer Based Information Systems in Primary Health Centers

    Science.gov (United States)

    Hosizah; Kuntoro; Basuki N., Hari

    2016-01-01

    The computer-based information system (CBIS) is adopted by almost all of in health care setting, including the primary health center in East Java Province Indonesia. Some of softwares available were SIMPUS, SIMPUSTRONIK, SIKDA Generik, e-puskesmas. Unfortunately they were most of the primary health center did not successfully implemented. This…

  18. Belgian experience in applying the {open_quotes}leak-before-break{close_quotes} concept to the primary loop piping

    Energy Technology Data Exchange (ETDEWEB)

    Gerard, R.; Malekian, C.; Meessen, O. [Tractebel Energy Engineering, Brussels (Belgium)

    1997-04-01

    The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports in the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.

  19. Innate immune cells in the pathogenesis of primary systemic vasculitis.

    Science.gov (United States)

    Misra, Durga Prasanna; Agarwal, Vikas

    2016-02-01

    Innate immune system forms the first line of defense against foreign substances. Neutrophils, eosinophils, erythrocytes, platelets, monocytes, macrophages, dendritic cells, γδ T cells, natural killer and natural killer T cells comprise the innate immune system. Genetic polymorphisms influencing the activation of innate immune cells predispose to development of vasculitis and influence its severity. Abnormally activated innate immune cells cross-talk with other cells of the innate immune system, present antigens more efficiently and activate T and B lymphocytes and cause tissue destruction via cell-mediated cytotoxicity and release of pro-inflammatory cytokines. These secreted cytokines further recruit other cells to the sites of vascular injury. They are involved in both the initiation as well as the perpetuation of vasculitis. Evidences suggest reversal of aberrant activation of immune cells in response to therapy. Understanding the role of innate immune cells in vasculitis helps understand the potential of therapeutic modulation of their activation to treat vasculitis.

  20. Progress in systemic chemotherapy of primary breast cancer: an overview.

    Science.gov (United States)

    Hortobagyi, G N

    2001-01-01

    Substantial progress has been made in the multidisciplinary management of primary breast cancer during the last 30 years. Adjuvant chemotherapy has been shown to significantly reduce the annual risk of cancer recurrence and mortality, and these effects persist even 15 years after diagnosis. Combination chemotherapy is superior to single-agent therapy and anthracycline-containing regimens. Those that combine an anthracycline with 5-fluorouracil and cyclophosphamide are more effective than regimens without an anthracycline. Six cycles of a single regimen appear to provide optimal benefit. Dose reductions below the standard range are associated with inferior results. Dose increases that require growth factor or hematopoietic stem cell support are under investigation; at this time, the existing results provide no compelling reason to use this strategy outside a clinical trial. Regimens using fixed crossover designs with two non-cross-resistant regimens are being evaluated. The addition of a taxane to anthracycline-containing regimens is currently under intense scrutiny, and preliminary analysis of the first three clinical trials has shown encouraging, albeit not compelling, results. For patients with estrogen receptor-positive breast cancer, the sequential administration of chemotherapy and 5 years of tamoxifen therapy provides additive benefits. No compelling evidence exists to combine ovarian ablation with chemotherapy. Most side effects and toxic effects are self-limited, although premature menopause requires monitoring and preventive interventions to preserve bone mineral density. The small risk of acute leukemia is of concern, and additional research to develop safer regimens is clearly indicated. The overall effect of optimal local/regional treatment combined with an anthracycline-containing adjuvant chemotherapy and a taxane (and, for patients with estrogen receptor-positive tumors, 5 years of tamoxifen therapy) is a greater than 50% reduction in annual risks of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  2. ANALISIS MODEL TERAS 3-DIMENSI UNTUK EVALUASI PARAMETER KRITIKALITAS REAKTOR PWR MAJU KELAS 1000 MW

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-04-01

    Full Text Available Setelah kejadian Fukushima, penggunaan sistem keselamatan pasif menjadi persyaratan yang penting untuk PLTN. PLTN jenis PWR maju kelas 1000 yang didesain oleh Westinghouse, AP1000, memiliki fitur keselamatan pasif disamping sederhana dan modular. Sebelum memilih suatu PLTN, maka perlu dilakukan suatu evaluasi terhadap parameter desainnya. Salah satu parameter yang penting dalam keselamatan adalah kritikalitas teras. Permasalahan pokok dalam mengevaluasi parameter kritikalitas teras AP1000 tidak adanya data komposisi material SS304 dan H2O di daerah reflektor dan diameter penyerap SS304. Dengan demikian tujuan penelitian ini adalah mendapatkan model teras 3-dimensi AP1000 dan siap diaplikasikan dalam evaluasi parameter kritikalitas teras. Hasil perhitungan menunjukkan bahwa komposisi terbaik SS304 dan H2O di reflektor teras bagian atas dan bawah masing-masing 50 vol%, sedangkan diameter penyerap SS304 adalah 0,960 cm. Evaluasi konsentrasi boron kritis menunjukkan perbedaan yang signifikan dengan nilai desain. Meskipun penyebab utama dari perbedaan ini belum diketahui, akan tetapi dapat dibuktikan bahwa konsentrasi boron kritis sangat sensitif dengan densitas UO2. Untuk reaktivitas padam, reaktor AP1000 memiliki margin subkritikalitas teras yang besar untuk satu siklus operasi. Dengan demikian teras yang diusulkan dapat digunakan sebagai acuan untuk evaluasi parameter teras lainnya atau perangkat analitis lainnya dalam rangka mengevaluasi desain reaktor AP1000. Kata kunci: AP1000, kritikalitas, konsentrasi boron kritis, reaktivitas padam   After the Fukushima accident, the use of passive safety system becomes an important requirement for the nuclear power plant (NPP. The advanced PWR NPP with 1000 MW (electric class, designed by Westinghouse, AP1000, a reactor with the passive safety features as well as simple and modular. Before selecting a nuclear power plant, there should be an evaluation of the design parameter. One important parameter in

  3. Primary proteolysis of white brined goat cheese monitored by high molarity Tris buffer SDS- PAGE system

    National Research Council Canada - National Science Library

    Milenko Smiljanić; Mirjana B. Pešić; Miroljub B. Barać; Sladjana P. Stanojević; Snežana T. Jovanović; Ognjen D. Maćej

    2013-01-01

    The aim of this work was to investigate primary proteolysis of white brined goat cheese prepared from raw milk and to correlate with the results obtained with high-molarity Tris buffer electrophoretic system...

  4. The development process for the space shuttle primary avionics software system

    Science.gov (United States)

    Keller, T. W.

    1987-01-01

    Primary avionics software system; software development approach; user support and problem diagnosis; software releases and configuration; quality/productivity programs; and software development/production facilities are addressed. Also examined are the external evaluations of the IBM process.

  5. Primary Health Care Software-A Computer Based Data Management System

    Directory of Open Access Journals (Sweden)

    Tuli K

    1990-01-01

    Full Text Available Realising the duplication and time consumption in the usual manual system of data collection necessitated experimentation with computer based management system for primary health care in the primary health centers. The details of the population as available in the existing manual system were used for computerizing the data. Software was designed for data entry and analysis. It was written in Dbase III plus language. It was so designed that a person with no knowledge about computer could use it, A cost analysis was done and the computer system was found more cost effective than the usual manual system.

  6. The Power-weakness Ratios (PWR as a Journal Indicator: Testing the “Tournaments” Metaphor in Citation Impact Studies

    Directory of Open Access Journals (Sweden)

    Loet Leydesdorff

    2016-09-01

    Full Text Available Purpose: Ramanujacharyulu developed the Power-weakness Ratio (PWR for scoring tournaments. The PWR algorithm has been advocated (and used for measuring the impact of journals. We show how such a newly proposed indicator can empirically be tested. Design/methodology/approach: PWR values can be found by recursively multiplying the citation matrix by itself until convergence is reached in both the cited and citing dimensions; the quotient of these two values is defined as PWR. We study the effectiveness of PWR using journal ecosystems drawn from the Library and Information Science (LIS set of the Web of Science (83 journals as an example. Pajek is used to compute PWRs for the full set, and Excel for the computation in the case of the two smaller sub-graphs: (1 JASIST+ the seven journals that cite JASIST more than 100 times in 2012; and (2 MIS Quart+ the nine journals citing this journal to the same extent. Findings: A test using the set of 83 journals converged, but did not provide interpretable results. Further decomposition of this set into homogeneous sub-graphs shows that—like most other journal indicators—PWR can perhaps be used within homogeneous sets, but not across citation communities. We conclude that PWR does not work as a journal impact indicator; journal impact, for example, is not a tournament. Research limitations: Journals that are not represented on the “citing” dimension of the matrix—for example, because they no longer appear, but are still registered as “cited” (e.g. ARIST—distort the PWR ranking because of zeros or very low values in the denominator. Practical implications: The association of “cited” with “power” and “citing” with “weakness” can be considered as a metaphor. In our opinion, referencing is an actor category and can be Metaphor in Citation Impact Studies in terms of behavior, whereas “citedness” is a property of a document with an expected dynamics very different from that of

  7. Venous thromboembolism in systemic autoimmune diseases: A narrative review with emphasis on primary systemic vasculitides.

    Science.gov (United States)

    Tamaki, Hiromichi; Khasnis, Atul

    2015-08-01

    Venous thromboembolism (VTE) is a prevalent multifactorial health condition associated with significant morbidity and mortality. Population-based epidemiological studies have revealed an association between systemic autoimmune diseases and deep venous thrombosis (DVT)/VTE. The etiopathogenesis of increased risk of VTE in systemic autoimmune diseases is not entirely clear but multiple contributors have been explored, especially in the context of systemic inflammation and disordered thrombogenesis. Epidemiologic data on increased risk of VTE in patients with primary systemic vasculitides (PSV) have accumulated in recent years and some of these studies suggest the increased risk while patients have active diseases. This could lead us to hypothesize that venous vascular inflammation has a role to play in this phenomenon, but this is unproven. The role of immunosuppressive agents in modulating the risk of VTE in patients with PSV is not yet clear except for Behçet's disease, where most of the studies are retrospective. Sensitizing physicians to this complication has implications for prevention and optimal management of patients with these complex diseases. This review will focus on the epidemiology and available evidence regarding pathogenesis, and will attempt to summarize the best available data regarding evaluation and treatment of these patients.

  8. Measurement of cold challenge responses in primary Raynaud's phenomenon and Raynaud's phenomenon associated with systemic sclerosis.

    OpenAIRE

    O'Reilly, D; Taylor, L.; el-Hadidy, K; Jayson, M I

    1992-01-01

    Using computed thermography continuous temperature recordings were made before and after cold challenge of the fingers of control subjects and patients with primary Raynaud's phenomenon and Raynaud's phenomenon associated with systemic sclerosis. Basal skin temperature measurements (Tpre) were significantly lower in patients with primary Raynaud's phenomenon and Raynaud's phenomenon associated with systemic sclerosis than in the controls. Temperatures immediately after cold challenge (T0) wer...

  9. Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, V.; Collins, B.L.; Sackett, K.E.; Coppin, C.E.

    1978-02-01

    Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, respectively, to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of these tests was to refine the definition of the upper limit of steam generator tube ruptures at which high peak cladding temperatures occur, as set by Test S-28-1. During these tests, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant in a PWR. Thirty (Test S-28-7), 34 (Test S-28-9), and 20 (Test S-28-12) steam generator tube ruptures were simulated by a controlled injection from a heated accmulator into the intact loop hot leg.

  10. Collaborative Decision Support Systems for Primary Health care Managers

    Directory of Open Access Journals (Sweden)

    Gunjan Pahuja

    2012-03-01

    Full Text Available In this paper, a collaborative DSS Model for health care systems and results obtained are described. The proposed framework [1] embeds expert knowledge within DSS to provide intelligent decision support, and implements the intelligent DSS using collaboration technologies. The problem space contains several Hub and Spoke networks. Information about such networks is dynamically captured and represented in a Meta-data table. This master table enables collaboration between any two networks in the problem space, through load transfer, between them. In order to show the collaboration the sample database of 15 health care centers is taken assuming that there are 5 health care centers in one network.

  11. Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V. [JSC ' Atomtechenergo' , Filial ' Novovoronezhatomtechenergo' , Novovorenezh (Russian Federation); Ryasny, Sergei I. [JSC ' Atomtechenergo' , Mytishi (Russian Federation)

    2016-09-15

    At WWER-1000 NPPs, as well as at PWR NPPs, there is a problem of determining the correct weighted mean coolant temperature in the primary circuit hot legs based on the measuring channels information. The problem is caused by the coolant temperature stratification. The technical documentation for engineering support and maintenance of I and C systems does not provide any regulatory guidelines to consider this effect. Therefore, it is very important to represent a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of the WWER-1000 reactor plants. The given paper presents the basic preconditions and approaches applied during the methodology development. They were worked out on the basis of the executed numerical and experimental research taking into account the analysis of the extensive material obtained by the authors from full-scale tests during the commissioning of WWER-1000 power units, as well as operational data obtained from several power units with different fuel loadings.

  12. Space Shuttle Program Primary Avionics Software System (PASS) Success Legacy - Quality and Reliability Date

    Science.gov (United States)

    Orr, James K.; Peltier, Daryl

    2010-01-01

    Thsi slide presentation reviews the avionics software system on board the space shuttle, with particular emphasis on the quality and reliability. The Primary Avionics Software System (PASS) provides automatic and fly-by-wire control of critical shuttle systems which executes in redundant computers. Charts given show the number of space shuttle flights vs time, PASS's development history, and other charts that point to the reliability of the system's development. The reliability of the system is also compared to predicted reliability.

  13. Modulation of Tumor Tolerance in Primary Central Nervous System Malignancies

    Directory of Open Access Journals (Sweden)

    Theodore S. Johnson

    2012-01-01

    Full Text Available Central nervous system tumors take advantage of the unique immunology of the CNS and develop exquisitely complex stromal networks that promote growth despite the presence of antigen-presenting cells and tumor-infiltrating lymphocytes. It is precisely this immunological paradox that is essential to the survival of the tumor. We review the evidence for functional CNS immune privilege and the impact it has on tumor tolerance. In this paper, we place an emphasis on the role of tumor-infiltrating myeloid cells in maintaining stromal and vascular quiescence, and we underscore the importance of indoleamine 2,3-dioxygenase activity as a myeloid-driven tumor tolerance mechanism. Much remains to be discovered regarding the tolerogenic mechanisms by which CNS tumors avoid immune clearance. Thus, it is an open question whether tumor tolerance in the brain is fundamentally different from that of peripheral sites of tumorigenesis or whether it simply stands as a particularly strong example of such tolerance.

  14. Metabolic Profiling of Systemic Lupus Erythematosus and Comparison with Primary Sjogren's Syndrome and Systemic Sclerosis.

    Directory of Open Access Journals (Sweden)

    Anders A Bengtsson

    Full Text Available Systemic lupus erythematosus (SLE is a chronic inflammatory autoimmune disease which can affect most organ systems including skin, joints and the kidney. Clinically, SLE is a heterogeneous disease and shares features of several other rheumatic diseases, in particular primary Sjögrens syndrome (pSS and systemic sclerosis (SSc, why it is difficult to diagnose The pathogenesis of SLE is not completely understood, partly due to the heterogeneity of the disease. This study demonstrates that metabolomics can be used as a tool for improved diagnosis of SLE compared to other similar autoimmune diseases. We observed differences in metabolic profiles with a classification specificity above 67% in the comparison of SLE with pSS, SSc and a matched group of healthy individuals. Selected metabolites were also significantly different between studied diseases. Biochemical pathway analysis was conducted to gain understanding of underlying pathways involved in the SLE pathogenesis. We found an increased oxidative activity in SLE, supported by increased xanthine oxidase activity and an increased turnover in the urea cycle. The most discriminatory metabolite observed was tryptophan, with decreased levels in SLE patients compared to control groups. Changes of tryptophan levels were related to changes in the activity of the aromatic amino acid decarboxylase (AADC and/or to activation of the kynurenine pathway.

  15. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  16. Coordinated Fast Primary Frequency Control from Offshore Wind Power Plants in MTDC System

    DEFF Research Database (Denmark)

    Sakamuri, Jayachandra N.; Hansen, Anca Daniela; Cutululis, Nicolaos Antonio

    2016-01-01

    In this paper, coordinated fast primary frequency control (FPFC) from offshore wind power plants (OWPPs) integrated to surrounding onshore AC power system through a three terminal VSC HVDC system is presented. The onshore AC grid frequency variations are emulated at offshore AC grid through...... and the dynamics of wind turbine are also discussed. The corresponding impact of OWPPs active power output variation at different wind speeds on the power system frequency control and DC grid voltage is also presented. The results show that the proposed coordinated fast primary frequency control from OWPPs...... improves the power system frequency while relieving the stress on the other AC grid participating in frequency control....

  17. Primary study of muscone's effect on cardiovascular system

    Institute of Scientific and Technical Information of China (English)

    ZHU Xue-jing; WU Qi-biao; LI Hai-tao

    2008-01-01

    Objective To investigate the effect of muscone on cardiovascular system. Methods Experimental animals to divide muscone high、middle、low dose group(the mouse is 20 mg·kg-1, 10 mg·kg-1, 5.0 mg·kg-1; the rat is 10 mg·kg-1, 5.0 mg·kg-1, 2.5 mg·kg-1), GT group( the mouse is 1/12 mg·kg-1; the rat is 1/24 mg·kg-1) and NS group. Intragastrie administration in a week, do the mouse ant-hypoxia experiment,the drug (Pit.) produce the rat myocardial ischemia experiment and obstruct coronary artery to produce the rat myocardial ischemia experiment. The mice's survival time (t), the rat's variation of T in electrocardiogram, creatinkinase (CK) and lactate dehydrogenase (LDH) were recorded, respectively. Results The effect of Muscone is significant difference between GT and NS in a dose variation manner. Conclusions Muscone has the effect of ant-hypoxia, cutting down T peak value, reducing CK and LDH. The muscone has effect to inhibiting myocardial ischemia.

  18. Study on stress corrosion of the zone affected by the AISI 316L steel heat under PWR reactor environment at 325 deg Celsius; Estudo da corrosao sob tensao da zona afetada pelo calor do aco AISI 316L em ambiente de reator PWR a 325 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Satler Filho, Luiz F.; Schvartzman, Monica M.A.M.; Quinan, Marco A.D.; Soares, Antonio E.G., E-mail: aegs@cdtn.b, E-mail: fernandosatler@yahoo.com.b, E-mail: quinanm@cdtn.b, E-mail: monicas@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Lima, Luciana I.L., E-mail: lill@cdtn.b [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2009-07-01

    This paper evaluates the stress corrosion susceptibility of the HAZ (heat affected zone) of the AISI 316L stainless steel of a dissimilar welding done between the ASTM A-508 steel and the AISI 316L steel, using a nickel alloy, under a chemical environment similar to the PWR (Pressurized Water Reactor) nuclear reactor primary circuit. The nickel 82 and 182 alloys were used in the GTAW (Gas Tungsten Arc Welding) and SMAW (Shielded Metal Arc Welding) processes respectively. The test at slow deformation - SSRT (Slow Strain Rate Test) was applied, using a deformation rate of 3x10{sup -7} s{sup -1}, at a temperature of 325 degree Celsius and pressure of 12.5 MPa. The susceptibility under tress corrosion evaluation was performed comparing the resistance limit, the total deformation and the fracture time obtained at the inert medium (nitrogen) and at the PWR medium. Also, the fracture surfaces were observed under a scanning electron microscope, verifying the fragile fracture regions

  19. 40 CFR 141.544 - What if my system uses chloramines, ozone, or chlorine dioxide for primary disinfection?

    Science.gov (United States)

    2010-07-01

    ..., ozone, or chlorine dioxide for primary disinfection? 141.544 Section 141.544 Protection of Environment... Benchmark § 141.544 What if my system uses chloramines, ozone, or chlorine dioxide for primary disinfection? If your system uses chloramines, ozone or chlorine dioxide for primary disinfection your system must...

  20. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  1. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  2. Patient-perceived responsiveness of primary care systems across Europe and the relationship with the health expenditure and remuneration systems of primary care doctors.

    Science.gov (United States)

    Murante, Anna Maria; Seghieri, Chiara; Vainieri, Milena; Schäfer, Willemijn L A

    2017-08-01

    Health systems are expected to be responsive, that is to provide services that are user-oriented and respectful of people. Several surveys have tried to measure all or some of the dimensions of the responsiveness (e.g. autonomy, choice, clarity of communication, confidentiality, dignity, prompt attention, quality of basic amenities, and access to family and community support), however there is little evidence regarding the level of responsiveness of primary care (PC) systems. This work analyses the capacity of primary care systems to be responsive. Data collected from 32 PC systems were used to investigate whether a relationship exists between the responsiveness of PC systems and the PC doctor remuneration systems and domestic health expenditure. There appears to be a higher responsiveness of PC when doctors are paid via capitation than when they only receive a fee for services or a mixed payment method. In addition, countries that spend more on health services are associated with higher levels of dignity and autonomy. Quality, as measured from the patient's perspective, does not necessarily overlap with PC performance based on structure and process indicators. The results could also stimulate a new debate on the role of economic resources and PC workforce payment mechanisms in the achievement of quality goals, in this case related to the capacity of PC systems to be responsive. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

  3. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  4. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  5. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  6. PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-10-01

    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

  7. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)

    2013-07-01

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  8. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  9. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  10. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  11. Colloids: a review of current knowledge with a view to application to phenomena of transportation within PWR; Colloides: point de vue sur les connaissances actuelles en vue d`une application aux phenomenes de transport dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Guinard, L.

    1996-12-31

    In an attempt to minimise dosimetry within the primary circuit of PWR units, research is being carried out into understanding the phenomena of transportation and deposition of corrosion products. It is therefore desirable to known the form of these corrosion products and the laws governing this form. It is generally considered that they are in soluble or particulate form. A third starts with a general presentation of colloids and goes on to define points which are useful, both on a theoretical and experimental level, in terms of application to phenomena of transportation within PWRs. (author). 69 refs., 30 figs., 6 tabs., 3 appends.

  12. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  13. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  14. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  15. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Energy Technology Data Exchange (ETDEWEB)

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  16. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  17. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Energy Technology Data Exchange (ETDEWEB)

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)

    2010-07-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  18. Microtensile bond strength of sealant and adhesive systems applied to occlusal primary enamel

    OpenAIRE

    Ramires-Romito, ACD; Reis, A; Loguercio, AD; Hipolito, VD; de Goes, MF; Singer, JD; Grande, RHM

    2007-01-01

    Purpose: To compare the microtensile bond strength (mu TBS) of a self-etch adhesive system (OptiBond Solo Self-Etch Adhesive System), two total etch adhesive system (OptiBond FL; OptiBond Solo), and a conventional sealant (Clinpro) applied to the occlusal surface of primary molars under saliva contamination. Methods: Sealant and adhesive systems were applied under manufacturers' specifications with or without previous saliva contamination. After storage in distilled water at 37 degrees C for ...

  19. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  20. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  1. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    Energy Technology Data Exchange (ETDEWEB)

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  2. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Science.gov (United States)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  3. Central nervous system involvement in primary Sjogren`s syndrome manifesting as multiple sclerosis.

    Science.gov (United States)

    Liu, Jing-Yao; Zhao, Teng; Zhou, Chun-Kui

    2014-04-01

    Central nervous system symptoms in patients with primary Sjogren`s syndrome are rare. They can present as extraglandular manifestations and require a differential diagnosis from multiple sclerosis. Due to a variety of presentations, Sjogren`s syndrome with neurologic involvement may be difficult to diagnose. Here, we report a case of a 75-year-old woman who was first diagnosed with multiple sclerosis in 2010, but who was subsequently diagnosed with primary Sjogren`s syndrome 2 years later after showing signs of atypical neurologic manifestations. Therefore, primary Sjogren`s syndrome should be suspected in patients who present with atypical clinical and radiologic neurologic manifestations.

  4. Sizing of an Energy Storage System for Grid Inertial Response and Primary Frequency Reserve

    DEFF Research Database (Denmark)

    Knap, Vaclav; Chaudhary, Sanjay Kumar; Stroe, Daniel Loan

    2016-01-01

    event in the power system with a high penetration of wind power. An energy storage system (ESS) might be a viable solution for providing inertial response and primary frequency regulation. A methodology has been presented here for the sizing of the ESS in terms of required power and energy. It describes...

  5. Gender Differences in the Primary Representational System according to Neurolinguistic Programming.

    Science.gov (United States)

    Cassiere, M. F.; And Others

    Neurolinguistic Programming (NLP) is a currently popular therapeutic modality in which individuals organize information through three basic sensory systems, one of which is the Primary Representational System (PRS). This study was designed to investigate gender differences in PRS according to the predicate preference method. It was expected that…

  6. Validation of the prognostic value of histologic scoring systems in primary sclerosing cholangitis

    DEFF Research Database (Denmark)

    de Vries, Elisabeth M G; de Krijger, Manon; Färkkilä, Martti

    2017-01-01

    Histologic scoring systems specific for primary sclerosing cholangitis (PSC) are not validated. We recently determined the applicability and prognostic value of three histological scoring systems in a single PSC cohort. The aim of this study was to validate their prognostic use and reproducibility...

  7. Influence of Micro-Grid in Steady State Performance of Primary Distribution System

    Directory of Open Access Journals (Sweden)

    K. Buayai

    2013-06-01

    Full Text Available Steady state analysis of primary distribution system is an integral part of Micro Grid (MG planning, design and operation of distribution system. In order to maximize performance and ensure secured operation of distribution system with MG, it is important to perform various analytical studies, both in static and dynamic domains. Static studies are the first step and static performance can be established by looking at a number of stead state aspects such as total power losses, voltage profile, feeder current and load ability of the system. This study presents such first step static analytical studies based on distribution load flow to see various steady state performances of primary distribution system due to the integration of MG. A 33-bus test distribution system has been used to present steady state performances. Results clearly show some useful contribution of MG in improving distribution system performance.

  8. Optimal Multi-type DGs Placement in Primary Distribution System by NSGA-II

    Directory of Open Access Journals (Sweden)

    K. Buayai

    2012-09-01

    Full Text Available The study proposes a multiobjective optimal placement of multi-type DG for enhancement of primary distribution system performance. A Pareto-based non-dominated sorting genetic algorithm II (NSGAII is proposed to determine locations and sizes of specified number of Distributed Generator units (DG within the primary distribution system. Three objective functions are considered as the indexes of the system performance: average Load Voltage Deviation (LVD minimization of the system real power loss and minimization of the annualized investment costs of DG. A fuzzy decision making analysis is used to obtain the final trade off optimal solution. The proposed methodology is tested on modified IEEE 33-bus radial system. Test results indicate that NSGA-II is a viable planning tool for practical DG placement and useful contribution of DG in improving the steady state system performance of the distribution system by the optimal allocation, setting and sizing multi-type DG.

  9. Hygrometric measurement for on-line monitoring of PWR vessel head penetrations; Detection de fuites de traversees de couvercles de cuve par surveillance hygrometrique

    Energy Technology Data Exchange (ETDEWEB)

    Germain, J.L.; Loisy, F.; Apolzan, S.

    1994-06-01

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

  10. Thermodynamics of greenhouse systems for the northern latitudes: analysis, evaluation and prospects for primary energy saving.

    Science.gov (United States)

    Bronchart, Filip; De Paepe, Michel; Dewulf, Jo; Schrevens, Eddie; Demeyer, Peter

    2013-04-15

    In Flanders and the Netherlands greenhouse production systems produce economically important quantities of vegetables, fruit and ornamentals. Indoor environmental control has resulted in high primary energy use. Until now, the research on saving primary energy in greenhouse systems has been mainly based on analysis of energy balances. However, according to the thermodynamic theory, an analysis based on the concept of exergy (free energy) and energy can result in new insights and primary energy savings. Therefore in this paper, we analyse the exergy and energy of various processes, inputs and outputs of a general greenhouse system. Also a total system analysis is then performed by linking the exergy analysis with a dynamic greenhouse climate growth simulation model. The exergy analysis indicates that some processes ("Sources") lie at the origin of several other processes, both destroying the exergy of primary energy inputs. The exergy destruction of these Sources is caused primarily by heat and vapour loss. Their impact can be compensated by exergy input from heating, solar radiation, or both. If the exergy destruction of these Sources is reduced, the necessary compensation can also be reduced. This can be accomplished through insulating the greenhouse and making the building more airtight. Other necessary Sources, namely transpiration and loss of CO2, have a low exergy destruction compared to the other Sources. They are therefore the best candidate for "pump" technologies ("vapour heat pump" and "CO2 pump") designed to have a low primary energy use. The combination of these proposed technologies results in an exergy efficient greenhouse with the highest primary energy savings. It can be concluded that exergy analyses add additional information compared to only energy analyses and it supports the development of primary energy efficient greenhouse systems.

  11. Development of Calculation Code for Fission Product and Corrosion Product in PWR’s Primary Loop

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; WAN; Hai-xia; SHAO; Jing; WU; Xiao-chun; LI; Long; LIU; Xing-min; KE; Guo-tu

    2015-01-01

    With the basis of study on generation,release and migration of fission product,calculation model for each of the above processes was developed,and calculation method for source term of PWR fission products was established.Study on source term of corrosion product in primary loop was been done.Based on the study of corrosion,

  12. Cognitive Multiple Access Network with Outage Margin in the Primary System

    DEFF Research Database (Denmark)

    Maham, Behrouz; Popovski, Petar; Zhou, Xiangyun

    2011-01-01

    This paper investigates the problem of spectrally efficient operation of a multiuser uplink cognitive radio system in the presence of a single primary link. The secondary system applies opportunistic interference cancelation (OIC) and decodes the primary signal when such an opportunity is created...... constraint at the primary system. We formulate the power optimization problem in various scenarios depending on the availability of channel state information and the type of power constraints, and propose a set of simple solutions. Finally, the analytical results are confirmed by simulations, indicating both...... the accuracy of the analysis, and the fact that the spectral-efficient, low-complexity, flexible, and high-performing cognitive radio can be designed based on the proposed schemes....

  13. [Primary malignant melanoma of the central nervous system: A diagnostic challenge].

    Science.gov (United States)

    Quillo-Olvera, Javier; Uribe-Olalde, Juan Salvador; Alcántara-Gómez, Leopoldo Alberto; Rejón-Pérez, Jorge Dax; Palomera-Gómez, Héctor Guillermo

    2015-01-01

    The rare incidence of primary malignant melanoma of the central nervous system and its ability to mimic other melanocytic tumors on images makes it a diagnostic challenge for the neurosurgeon. A 51-year-old patient, with a tumor located in the right forniceal callosum area. Total surgical excision was performed. Histopathological result was consistent with the diagnosis of primary malignant melanoma of the central nervous system, after ruling out extra cranial and extra spinal melanocytic lesions. The primary malignant melanoma of the central nervous system is extremely rare. There are features in magnetic resonance imaging that increase the diagnostic suspicion; nevertheless there are other tumors with more prevalence that share some of these features through image. Since there is not an established therapeutic standard its prognosis is discouraging. Copyright © 2015 Academia Mexicana de Cirugía A.C. Published by Masson Doyma México S.A. All rights reserved.

  14. Europe's strong primary care systems are linked to better population health but also to higher health spending.

    NARCIS (Netherlands)

    Kringos, D.S.; Boerma, W.; Zee, J. van der; Groenewegen, P.

    2013-01-01

    Strong primary care systems are often viewed as the bedrock of health care systems that provide high-quality care, but the evidence supporting this view is somewhat limited. We analyzed comparative primary care data collected in 2009-10 as part of a European Union-funded project, the Primary Health

  15. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  16. A human factors systems approach to understanding team-based primary care: a qualitative analysis.

    Science.gov (United States)

    Mundt, Marlon P; Swedlund, Matthew P

    2016-12-01

    Research shows that high-functioning teams improve patient outcomes in primary care. However, there is no consensus on a conceptual model of team-based primary care that can be used to guide measurement and performance evaluation of teams. To qualitatively understand whether the Systems Engineering Initiative for Patient Safety (SEIPS) model could serve as a framework for creating and evaluating team-based primary care. We evaluated qualitative interview data from 19 clinicians and staff members from 6 primary care clinics associated with a large Midwestern university. All health care clinicians and staff in the study clinics completed a survey of their communication connections to team members. Social network analysis identified key informants for interviews by selecting the respondents with the highest frequency of communication ties as reported by their teammates. Semi-structured interviews focused on communication patterns, team climate and teamwork. Themes derived from the interviews lent support to the SEIPS model components, such as the work system (Team, Tools and Technology, Physical Environment, Tasks and Organization), team processes and team outcomes. Our qualitative data support the SEIPS model as a promising conceptual framework for creating and evaluating primary care teams. Future studies of team-based care may benefit from using the SEIPS model to shift clinical practice to high functioning team-based primary care. © The Author 2016. Published by Oxford University Press. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  17. Comparison of carbon and HI-Z primary collimators for the LHC phase II collimation system

    CERN Document Server

    Keller, L; Smith, J; Assmann, R; Bracco, C; Weiler, T

    2010-01-01

    A current issue with the LHC collimation system is single-diffractive, off-energy protons from the primary collimators that pass completely through the secondary collimation system and are absorbed immediately downbeam in the cold magnets of the dispersion suppressor section. Simulations suggest that the high impact rate could result in quenching of these magnets. We have studied replacing the 60 cm primary graphite collimators, which remove halo mainly by inelastic strong interactions, with 5.25 mm tungsten, which remove halo mainly by multiple coulomb scattering and thereby reduce the rate of single-diffractive interactions that cause losses in the dispersion suppressor.

  18. Primary central nervous system lymphoma presenting as isolated oculomotor nerve palsy

    Directory of Open Access Journals (Sweden)

    Terence Tan, MBBS

    2014-09-01

    Full Text Available The authors report an unusual case of primary central nervous system lymphoma presenting with isolated pupil-involved oculomotor nerve palsy. Magnetic resonance imaging demonstrated leptomeningeal involvement of the midbrain and interpeduncular cistern, a single hypothalamic lesion, and intraventricular involvement. Diffuse large B-cell lymphoma was confirmed by stereotactic intraventricular biopsy. Combination chemotherapy with methotrexate, vincristine, procarbazine and rituximab was instituted with resolution of oculomotor nerve palsy and complete disease remission. An interdisciplinary approach involving neurosurgeons, neuroradiologists, neuropathologists and neurologists is crucial in the management of primary central nervous system lymphoma.

  19. Office systems for promoting screening mammography. A survey of primary care practices.

    Science.gov (United States)

    Melville, S K; Luckmann, R; Coghlin, J; Gann, P

    1993-12-01

    Office tracking, scheduling, and reminder systems have been shown to improve utilization of screening mammography, but little is known about the use of these systems by primary care physicians. We surveyed 132 primary care and obstetrics and gynecology practices affiliated with an independent practice association model health maintenance organization in central Massachusetts to determine their use of reminder, scheduling, and follow-up systems, and education and counseling services aimed at increasing screening mammography rates. The use of chart flags to remind physicians of a patient's need for mammography screening was reported by 30% of practices. Thirty-one percent reported the use of flow sheets, and 27% reported the use of mail or telephone patient reminders. At least one of these three systems was used by 57% of the practices, whereas 43% reported having none of these three systems. Variations in the use of these office systems were related to specialty type, physician number, and clinical staffing. The majority of practices (77%) reported using written educational materials, and 42% offered prevention counseling with nonphysician staff. Very few offices (8%) reported using mail or telephone reminders for previously scheduled appointments. Despite the proven effectiveness of reminder systems for screening mammography, many practices do not have a system in place. Promotion of reminder systems in primary care practices could have a substantial impact on mammography utilization.

  20. Importance of tip sensing for active control system of 30-m RIT primary mirror

    Institute of Scientific and Technical Information of China (English)

    Yichun Dai; Zhong Liu; Zhenyu Jin

    2009-01-01

    The active control of 30-m ring interferometric telescope(RIT)needs edge sensing and tip sensing when its primary mirror is composed by trapezoid-shaped segments,and the imaging performance of the RIT is determined by the accuracy of these two detecting approaches.Considering the detecting accuracy available in current segmented telescope active control systems,the effect of these detecting approaches on the surface error of the RIT primary mirror is calculated from the point of error propagation.The corresponding effect on imaging performance(modulation transfer functions(MTFs)and point spread functions(PSFs)at several typical wavelengths)of the RIT primary mirror is also simulated.The results show that tip sensing is very important for increasing the active control quality of the RIT primary mirror under the present techniques.

  1. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  2. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  3. [Differences and similarities of primary care in the German and Spanish health care systems].

    Science.gov (United States)

    Salvador Comino, María Rosa; Krane, Sibylla; Schelling, Jörg; Regife García, Víctor

    2016-02-01

    An efficient primary care is of particular importance for any countries' health care system. Many differences exist on how distinctive countries try to obtain the goal of an efficient, cost-effective primary care for its population. In this article we conducted a selective literature review, which includes both scientific and socio-political publications. The findings are complemented with the experience of a Spanish physician from Seville in her last year of training in family medicine, who completed a four months long rotation in the German health care system. We highlighted different features by comparing both countries, including their health care expenditure, the relation between primary and secondary care, the organization in the academic field and the training of future primary care physicians. It is clear that primary care in both countries plays a central role, have to deal with shortcomings, and in some points one system can learn from the other. Copyright © 2015 Elsevier España, S.L.U. All rights reserved.

  4. Primary Care Behavioral Health Provider Training: Systematic Development and Implementation in a Large Medical System.

    Science.gov (United States)

    Dobmeyer, Anne C; Hunter, Christopher L; Corso, Meghan L; Nielsen, Matthew K; Corso, Kent A; Polizzi, Nicholas C; Earles, Jay E

    2016-09-01

    The expansion of integrated, collaborative, behavioral health services in primary care requires a trained behavioral health workforce with specific competencies to deliver effective, evidence-informed, team-based care. Most behavioral health providers do not have training or experience working as primary care behavioral health consultants (BHCs), and require structured training to function effectively in this role. This article discusses one such training program developed to meet the needs of a large healthcare system initiating widespread implementation of the primary care behavioral health model of service delivery. It details the Department of Defense's experience in developing its extensive BHC training program, including challenges of addressing personnel selection and hiring issues, selecting a model for training, developing and implementing a phased training curriculum, and improving the training over time to address identified gaps. Future directions for training improvements and lessons learned in a large healthcare system are discussed.

  5. A continuous hyperspatial monitoring system of evapotranspiration and gross primary productivity from Unmanned Aerial Systems

    Science.gov (United States)

    Wang, Sheng; Bandini, Filippo; Jakobsen, Jakob; Zarco-Tejada, Pablo J.; Köppl, Christian Josef; Haugård Olesen, Daniel; Ibrom, Andreas; Bauer-Gottwein, Peter; Garcia, Monica

    2017-04-01

    Unmanned Aerial Systems (UAS) can collect optical and thermal hyperspatial (signatures of the land surface, closely linked with the vegetation structure and functioning, are already part of models to predict Evapotranspiration (ET) and Gross Primary Productivity (GPP) from satellites. However, there remain challenges for an operational monitoring using UAS compared to satellites: the payload capacity of most commercial UAS is less than 2 kg, but miniaturized sensors have low signal to noise ratios and small field of view requires mosaicking hundreds of images and accurate orthorectification. In addition, wind gusts and lower platform stability require appropriate geometric and radiometric corrections. Finally, modeling fluxes on days without images is still an issue for both satellite and UAS applications. This study focuses on designing an operational UAS-based monitoring system including payload design, sensor calibration, based on routine collection of optical and thermal images in a Danish willow field to perform a joint monitoring of ET and GPP dynamics over continuous time at daily time steps. The payload (digital camera (Sony RX-100) used to retrieve accurate digital elevation models (DEMs) for multispectral and thermal image orthorectification, and a standard GNSS single frequency receiver (UBlox) or a real time kinematic double frequency system (Novatel Inc. flexpack6+OEM628). Geometric calibration of the digital and multispectral cameras was conducted to recover intrinsic camera parameters. After geometric calibration, accurate DEMs with vertical errors about 10cm could be retrieved. Radiometric calibration for the multispectral camera was conducted with an integrating sphere (Labsphere CSTM-USS-2000C) and the laboratory calibration showed that the camera measured radiance had a bias within ±4.8%. The thermal camera was calibrated using a black body at varying target and ambient temperatures and resulted in laboratory accuracy with RMSE of 0.95 K. A

  6. Release and systemic accumulation of heavy metals from preformed crowns used in restoration of primary teeth.

    Science.gov (United States)

    Kodaira, Hiroe; Ohno, Kohachiro; Fukase, Naoko; Kuroda, Midori; Adachi, Shiki; Kikuchi, Motohiro; Asada, Yoshinobu

    2013-01-01

    Preformed crowns for restoration of primary teeth are used in various treatments and are essential for restoring the crowns of primary molars. However, there are concerns that mechanical, chemical, and thermal stimulation may cause release of components of such crowns. We examined systemic accumulation of heavy metals associated with preformed crowns (3M Stainless Steel Primary Molar Crowns) used in primary tooth restoration. The participants were 37 children who had visited the Pediatric Dental Clinic of Tsurumi University Dental Hospital. They were divided into two groups: 22 participants without a history of being fitted with a preformed crown for primary tooth restoration (controls), and 15 participants with preformed crowns for primary tooth restoration. Analysis of hair samples showed a significant difference in the level of the trace element Cr - an important component of the preformed crowns - between children with and without preformed crowns, but no significant differences in Fe or Ni levels. Levels of the trace elements Ni, Cr, and Fe were within allowable ranges, indicating that these minerals were not likely to be harmful.

  7. Mammaglobin-A immunohistochemistry in primary central nervous system neoplasms and intracranial metastatic breast carcinoma.

    Science.gov (United States)

    Cimino, Patrick J; Perrin, Richard J

    2014-07-01

    Metastases represent the most common type of intracranial neoplasm. In women, 30% of such tumors derive from breast carcinoma. In neurosurgical cases with ambiguous cellular morphology and/or limited biopsy material, immunohistochemistry (IHC) is often performed to distinguish metastases from primary central nervous system (CNS) neoplasms. IHC for mammaglobin-A (MGA), a protein expressed in a majority of breast carcinomas, is commonly applied in this setting, but its utility for distinguishing primary CNS neoplasms from metastatic breast carcinoma is unknown; the reactivity of MGA in primary and metastatic CNS neoplasms has never been described. Here, we describe the frequency and patterns of IHC reactivity for MGA in metastatic and primary CNS neoplasms from patients with well-documented histories of breast carcinoma. Following a published protocol previously applied to non-CNS neoplasms, MGA staining of moderate to strong intensity within 5% or more of a neoplasm was considered positive. On the basis of these criteria, 3 of 12 (25.0%) glioblastomas, 1 of 10 (10.0%) meningiomas, and 47 of 95 (49.5%) metastases were positive. Importantly, the cytoarchitectural staining characteristics among all 4 MGA-positive primary CNS neoplasms (cytoplasmic and nuclear) differed from those of the metastases (cytoplasmic and membranous). These findings suggest that MGA IHC staining intensity and distribution can distinguish metastases from primary CNS neoplasms (P=0.0086) in women with a history of breast carcinoma but also indicate that cytologic staining patterns must be interpreted for more accurate tumor classification.

  8. Development of Quality Assurance System in Culture and Nation Character Education in Primary Education in Indonesia

    Science.gov (United States)

    Susilana, Rudi; Asra

    2013-01-01

    The purpose of national education is to develop skills and build dignified national character and civilization in educating nation life (Act No. 20, 2003). The paper describes a system of quality assurance in culture and character education in primary education. This study employs the six sigma model which consists of the formula DMAIC (Define,…

  9. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Science.gov (United States)

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Regulatory Commission (NRC) is issuing Revision 2 to Regulatory Guide (RG) 1.57, ``Design Limits and Loading... clarity. This guide describes a method that the NRC staff considers acceptable for design limits...

  10. Assessing Primary Representational System (PRS) Preference for Neurolinguistic Programming (NLP) Using Three Methods.

    Science.gov (United States)

    Dorn, Fred J.

    1983-01-01

    Considered three methods of identifying Primary Representational System (PRS)--an interview, a word list, and a self-report--in a study of 120 college students. Results suggested the three methods offer little to counselors either collectively or individually. Results did not validate the PRS construct, suggesting the need for further research.…

  11. Assessment of Primary Representational Systems with Neurolinguistic Programming: Examination of Preliminary Literature.

    Science.gov (United States)

    Dorn, Fred J.; And Others

    1983-01-01

    Reviews the inconsistent findings of studies on neurolinguistic programing and recommends some areas that should be examined to verify various claims. Discusses methods of assessing client's primary representational systems, including predicate usage and eye movements, and suggests that more reliable methods of assessing PRS must be found. (JAC)

  12. Use of email in communication between the Finnish primary healthcare system and general practitioners

    Directory of Open Access Journals (Sweden)

    Tuula Karhula

    2011-03-01

    Conclusion Especially during periods of change in the workplace, it is very important that management is conducted personally. Care must be taken so that disinformation does not spoil the informative value of email in the administration of primary health care. The needed technical assistance should be given to everyone in order to get the best advantage from the use of the email system.

  13. Predictive Value of Tokuhashi Scoring Systems in Spinal Metastases, Focusing on Various Primary Tumor Groups

    DEFF Research Database (Denmark)

    Wang, Miao; Bünger, Cody; Haisheng, Li

    2012-01-01

    STUDY DESIGN: We conducted a prospective cohort study of 448 patients with spinal metastases from a variety of cancer groups. OBJECTIVE: To determine the specific predictive value of the Tokuhashi scoring system (T12) and its revised version (T15) in spinal metastases of various primary tumors. S...

  14. Complexity in practice: understanding primary care as a complex adaptive system

    Directory of Open Access Journals (Sweden)

    Beverley Ellis

    2010-06-01

    Conclusions The results are real-world exemplars of the emergent properties of complex adaptive systems. Improving clinical governance in primary care requires both complex social interactions and underpinning informatics. The socio-technical lessons learned from this research should inform future management approaches.

  15. Post-transplantation primary central nervous system lymphoma in a patient with systemic lupus erythematosus and prolonged use of immunosuppressant.

    Science.gov (United States)

    Tse, Teresa P K; Chan, Allan N L; Chan, Tony K T; Po, Y C

    2014-12-01

    Post-transplantation primary central nervous system lymphoma is an uncommon and fatal post-transplant lymphoproliferative disorder. Such lymphomas have been described in only a few case series in the literature. The incidence of this condition is rising with improved survival after organ transplantation. A case of post-transplantation primary central nervous system lymphoma in a young Chinese woman with systemic lupus erythematosus is described here. She presented with right-sided weakness and memory loss after tooth extraction 2 weeks before admission. Contrast computed tomography of the brain demonstrated a contrast rim-enhancing lesion over the left frontal lobe. With a history of recent dental procedure, long-term immunosuppressive therapy and computed tomography findings, cerebral abscess was highly suspected. Emergency operation was performed. Histopathology showed post-transplantation primary central nervous system lymphoma, with cells positive for B-cell marker CD20. Immunosuppressant was stopped and she was treated with radiotherapy and rituximab (anti-CD20 monoclonal antibody). She remained disease-free at 16 months. Post-transplantation primary central nervous system lymphoma is rare with variable presentation and radiological features. We believe rituximab may have a role in the treatment of such lymphomas.

  16. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  17. Investigating Power System Primary and Secondary Reserve Interaction under High Wind Power Penetration

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yingchen [National Renewable Energy Lab. (NREL), Golden, CO (United States); Tan, Jin [National Renewable Energy Lab. (NREL), Golden, CO (United States); Krad, Ibrahim [National Renewable Energy Lab. (NREL), Golden, CO (United States); Yang, Rui [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gevorgian, Vahan [National Renewable Energy Lab. (NREL), Golden, CO (United States); Ela, Erik [Electric Power Research Inst. (EPRI), Knoxville, TN (United States)

    2016-12-01

    Power system frequency needs to be maintained close to its nominal value at all times to successfully balance load and generation and maintain system reliability. Adequate primary frequency response and secondary frequency response are the primary forces to correct an energy imbalance at the second-to-minute level. As wind energy becomes a larger portion of the world's energy portfolio, there is an increased need for wind to provide frequency response. This paper addresses one of the major concerns about using wind for frequency regulation: the unknown factor of the interaction between primary and secondary reserves. The lack of a commercially available tool to model this has limited the energy industry's understanding of when the depletion of primary reserves will impact the performance of secondary response or vice versa. This paper investigates the issue by developing a multi-area frequency response integration tool with combined primary and secondary capabilities. The simulation is conducted in close coordination with economical energy scheduling scenarios to ensure credible simulation results.

  18. Implementation of IEEE 802.15.4a Based UWB Systems for Coexistence with Primary Users

    Directory of Open Access Journals (Sweden)

    Caglar Findikli

    2016-03-01

    Full Text Available Peaceful coexistence is a major implementation issuefor both cognitive radios and ultra wideband (UWB systems.Accordingly, the UWB impulse radio (UWB-IR based WirelessPersonal Area Network (WPAN standard IEEE 802.15.4a hassuggested using linear combination of pulse to limit interfer-ence to coexisting primary systems. In this paper, motivatedby implementing the IEEE 802.15.4a based UWB-IR systemsfor peaceful coexistence, we consider the implementation oflinear combination of pulses as suggested by the standard.Accordingly, we (i design possible linearly combined pulses thatconform to the standard requirements, (ii consider coherentand noncoherent receiver structures that can be adapted for thephysical layer of the IEEE 802.15.4a standard, (iii investigatethe effect of channel models on the system performance, and(iv study the UWB-IR system performance in the presenceof narrowband and orthogonal frequency division multiplexing(OFDM based wideband primary systems with various band-widths and subcarriers. The study shows that the UWB-IR systemperformance can be significantly improved by selecting suitablepulses for transmission and employing appropriate filteringtechniques at the receiver when the primary system is active.For the implementation of IEEE 802.15.4a based UWB systemscomplying with coexistence requirements, the results of this studyshould be carefully considered.

  19. Structural and contextual dimensions of Iranian primary health care system at local level.

    Science.gov (United States)

    Zanganeh Baygi, Mehdi; Seyedin, Hesam; Salehi, Masoud; Jafari Sirizi, Mehdi

    2015-01-01

    In recent years, family physician plan was established as the main strategy of health system in Iran, while organizational structure of the primary health care system has remained the same as thirty years ago. This study was performed to illustrate structural and contextual dimensions of organizational structure and relationship between them in Iranian primary health care system at local level. A cross-sectional quantitative study was conducted from January to June 2013, during which 121 questionnaires were distributed among senior and junior managers of city health centers at Medical Sciences universities in Iran. Validity of the questionnaire was confirmed by experts (CVI = 0.089 and CVR more than 0.85) and Cronbach α was utilized for reliability (α = 0.904). We used multistage sampling method in this study and analysis of the data was performed by SPSS software using different tests. Local level of primary health care system in Iran had mechanical structure, but in contextual dimensions the results showed different types. There was a significant relationship between structural and contextual dimensions (r = 0.642, P value structural dimensions. Because of the changes in goals and strategies of Iranian health system in recent years, it is urgently recommended to reform the current structure to increase efficiency and effectiveness of the system.

  20. [The primary research and development of software oversampling mapping system for electrocardiogram].

    Science.gov (United States)

    Zhou, Yu; Ren, Jie

    2011-04-01

    We put forward a new concept of software oversampling mapping system for electrocardiogram (ECG) to assist the research of the ECG inverse problem to improve the generality of mapping system and the quality of mapping signals. We then developed a conceptual system based on the traditional ECG detecting circuit, Labview and DAQ card produced by National Instruments, and at the same time combined the newly-developed oversampling method into the system. The results indicated that the system could map ECG signals accurately and the quality of the signals was good. The improvement of hardware and enhancement of software made the system suitable for mapping in different situations. So the primary development of the software for oversampling mapping system was successful and further research and development can make the system a powerful tool for researching ECG inverse problem.

  1. Micro-tensile bond strength of adhesive systems applied on occlusal primary enamel.

    Science.gov (United States)

    Ramires-Romito, Ana Cláudia; Reis, Alessandra; Loguercio, Alessandro Dourado; de Góes, Mario Fernando; Grande, Rosa Helena Miranda

    2004-01-01

    The aim of this study was to evaluate the micro-tensile bond strength of adhesive systems (OptiBond Solo, Kerr; Prime & Bond NT, Dentsply) on occlusal surface of primary molars. The adhesives were tested under manufacturers' specifications and after contamination of the bonding site with saliva. Hourglass cylindrical-shaped samples were obtained and subjected to a tensile force. No significant difference was observed among the groups. OptiBond Solo and Prime & Bond NT showed similar values of bond strengths when applied on occlusal enamel of primary molar under either saliva contamination or not.

  2. Comparing the Asthma APGAR system and the Asthma Control Test™ in a multicenter primary care sample.

    Science.gov (United States)

    Rank, Matthew A; Bertram, Susan; Wollan, Peter; Yawn, Roy A; Yawn, Barbara P

    2014-07-01

    To compare asthma control assessment using the Asthma APGAR system, a tool developed by primary care clinicians, in a multicenter primary care sample with the Asthma Control Test (ACT™)/Childhood Asthma Control Test (CACT™), a tool developed by asthma specialists. This is a substudy of a multicenter, randomized, controlled pragmatic trial that tests the effectiveness of the Asthma APGAR system in primary care practices. As part of the study, enrolled patients completed both the ACT™/CACT™ and the Asthma APGAR system between March 1, 2011, and December 31, 2011. Kappa and McNemar statistics were used to compare the results of questionnaires. Of the 468 patients in our sample, 306 (65%) were classified as not controlled by the ACT™/CACT™ or the Asthma APGAR system. The overall agreement was 84.4%, with a kappa value of .68 (substantial agreement) and a McNemar test P value of .35 (suggesting no significant difference in the direction of disagreement). Of those with poor control as defined by the Asthma APGAR system, 23.8% (73) had no controller medications and 76.5% (234) were seldom or sometimes able to avoid identified triggers for their asthma. Of those who stated that they had been prescribed controller medications, 116 of 332 (35%) stated that they did not use the controller medication on a daily basis. The Asthma APGAR system and the ACT™/CACT™ similarly assess asthma control in a multicenter primary care-based sample. The Asthma APGAR system identified an "actionable item" in more than 75% (234) of the individuals with poor asthma control, thus linking an assessment of poor asthma control with a management strategy. Copyright © 2014 Mayo Foundation for Medical Education and Research. Published by Elsevier Inc. All rights reserved.

  3. Getting on with your computer is associated with job satisfaction in primary care: entrants to primary care should be assessed for their competency with electronic patient record systems

    Directory of Open Access Journals (Sweden)

    Simon de Lusignan

    2014-02-01

    Full Text Available Job satisfaction in primary care is associated with getting on with your computer. Many primary care professionals spend longer interacting with their computer than anything else in their day. However, the computer often makes demands rather than be an aid or supporter that has learned its user’s preferences. The use of electronic patient record (EPR systems is underrepresented in the assessment of entrants to primary care, and in definitions of the core competencies of a family physician/general practitioner. We call for this to be put right: for the use of the EPR to support direct patient care and clinical governance to be given greater prominence in training and assessment. In parallel, policy makers should ensure that the EPR system use is orientated to ensuring patients receive evidence-based care, and EPR system suppliers should explore how their systems might better support their clinician users, in particular learning their preferences.

  4. Getting on with your computer is associated with job satisfaction in primary care: entrants to primary care should be assessed for their competency with electronic patient record systems.

    Science.gov (United States)

    de Lusignan, Simon; Pearce, Christopher; Munro, Neil

    2013-01-01

    Job satisfaction in primary care is associated with getting on with your computer. Many primary care professionals spend longer interacting with their computer than anything else in their day. However, the computer often makes demands rather than be an aid or supporter that has learned its user's preferences. The use of electronic patient record (EPR) systems is underrepresented in the assessment of entrants to primary care, and in definitions of the core competencies of a family physician/general practitioner. We call for this to be put right: for the use of the EPR to support direct patient care and clinical governance to be given greater prominence in training and assessment. In parallel, policy makers should ensure that the EPR system use is orientated to ensuring patients receive evidence-based care, and EPR system suppliers should explore how their systems might better support their clinician users, in particular learning their preferences.

  5. Factors militating against effective implementation of primary health care (PHC system in Nigeria

    Directory of Open Access Journals (Sweden)

    Josephat M Chinawa

    2015-01-01

    Full Text Available Objectives: This study aimed to evaluate the factors that militate against effective implementation of a primary health care (PHC system in Nigeria. Materials and Methods: The study was conducted at four selected PHC centers in Enugu State from November 2014 to January 2015. The primary health center was chosen by systemic sampling from about eight primary health centers in Enugu metropolis. The sixteen-item questionnaire was elaborated with the Likert scale. Data retrieved were collected with the aid of a structured study pro forma and analyzed using SPSS Version 18. Results: A total of 169 health workers were recruited from four primary health centers. The mean age of all participants was 38.42 years standard deviation (SD = 9.8, while the male: Female ratio was 2:1. Among the subjects, 59% were aged 30-39 years. Existing equipment and manpower on one hand and job security and salary on the other hand are negative factors in the implementation of PHC; the respondents believed that adequate supply of gloves, needles, bandages, good access to drugs and medications, a good cold chain system, and full implementation of immunization programs all exist in PHC centers. Adequate community participation, culture and religion, access to safe and clean water, and steady electricity, on the other hand, are nonexistent in the PHC centers in the study. Conclusions: The PHC centers studied showed that much remains to be desired, especially in terms of manpower, communication, and the remuneration of health workers.

  6. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  7. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  8. Payment systems and incentives in primary care: implications of recent reforms in Estonia and Romania.

    Science.gov (United States)

    Dan, Sorin; Savi, Riin

    2015-01-01

    Since the early 1990s, major reform in healthcare has been adopted in former communist countries in Central and Eastern Europe. More than 20 years after, reform in healthcare still draws much interest from policy makers and academics alike. One of the dynamic components of reform has been the reform of payment systems in primary care. This article looks at recent developments in payment systems and financial incentives in Estonia and Romania. We conclude that finding the appropriate mix in paying and incentivizing primary care providers in a transitional context is no easy solution for healthcare policy makers who need to carefully weigh in the advantages and inherent problems of various payment arrangements. In a transitional, rapidly changing healthcare system and society, and a context of financial stringency, the theoretical effects of payment mechanisms may be more difficult to predict and manage than it is expected.

  9. Equity in primary health care delivery: an examination of the cohesiveness of strategies relating to the primary healthcare system, the health workforce and hepatitis C.

    Science.gov (United States)

    Scarborough, Jane; Eliott, Jaklin; Miller, Emma; Aylward, Paul

    2015-04-01

    To suggest ways of increasing the cohesiveness of national primary healthcare strategies and hepatitis C strategies, with the aim of ensuring that all these strategies include ways to address barriers and facilitators to access to primary healthcare and equity for people with hepatitis C. A critical review was conducted of the first national Primary Healthcare System Strategy and Health Workforce Strategy with the concurrent Hepatitis C Strategy. Content relating to provision of healthcare in private general practice was examined, focussing on issues around access and equity. In all strategies, achieving access to care and equity was framed around providing sufficient medical practitioners for particular locations. Equity statements were present in all policies but only the Hepatitis C Strategy identified discrimination as a barrier to equity. Approaches detailed in the Primary Healthcare System Strategy and Health Workforce Strategy regarding current resource allocation, needs assessment and general practitioner incentives were limited to groups defined within these documents and may not identify or meet the needs of people with hepatitis C. Actions in the primary healthcare system and health workforce strategies should be extended to additional groups beyond those listed as priority groups within the strategies. Future hepatitis C strategies should outline appropriate, detailed needs assessment methodologies and specify how actions in the broad strategies can be applied to benefit the primary healthcare needs of people with hepatitis C.

  10. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  11. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  12. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  13. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  14. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  15. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  16. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  17. Integration of Lithium-Ion Battery Storage Systems in Hydroelectric Plants for Supplying Primary Control Reserve

    Directory of Open Access Journals (Sweden)

    Fabio Bignucolo

    2017-01-01

    Full Text Available The ever-growing diffusion of renewables as electrical generation sources is forcing the electrical power system to face new and challenging regulation problems to preserve grid stability. Among these, the primary control reserve is reckoned to be one of the most important issues, since the introduction of generators based on renewable energies and interconnected through static converters, if relieved from the primary reserve contribution, reduces both the system inertia and the available power reserve in case of network events involving frequency perturbations. In this scenario, renewable plants such as hydroelectric run-of-river generators could be required to provide the primary control reserve ancillary service. In this paper, the integration between a multi-unit run-of-river power plant and a lithium-ion based battery storage system is investigated, suitably accounting for the ancillary service characteristics as required by present grid codes. The storage system is studied in terms of maximum economic profitability, taking into account its operating constraints. Dynamic simulations are carried out within the DIgSILENT PowerFactory 2016 software environment in order to analyse the plant response in case of network frequency contingencies, comparing the pure hydroelectric plant with the hybrid one, in which the primary reserve is partially or completely supplied by the storage system. Results confirm that the battery storage system response to frequency perturbations is clearly faster and more accurate during the transient phase compared to a traditional plant, since time delays due to hydraulic and mechanical regulations are overpassed. A case study, based on data from an existing hydropower plant and referring to the Italian context in terms of operational constraints and ancillary service remuneration, is presented.

  18. Effects of Hospital Systems on Medical Home Transformation in Primary Care Residency Training Practices.

    Science.gov (United States)

    Knierim, Kyle; Hall, Tristen; Fernald, Douglas; Staff, Thomas J; Buscaj, Emilie; Allen, Jessica Cornett; Onysko, Mary; Dickinson, W Perry

    2016-11-23

    Most primary care residency training practices have close financial and administrative relationships with teaching hospitals and health systems. Many residency practices have begun integrating the core principles of the patient-centered medical home (PCMH) into clinical workflows and educational experiences. Little is known about how the relationships with hospitals and health systems affect these transformation efforts. Data from the Colorado Residency PCMH Project were analyzed. Results show that teaching hospitals and health systems have significant opportunities to influence residency practices' transformation, particularly in the areas of supporting team-based care, value-based payment reforms, and health information technology.

  19. Designing display primaries with currently available light sources for UHDTV wide-gamut system colorimetry.

    Science.gov (United States)

    Masaoka, Kenichiro; Nishida, Yukihiro; Sugawara, Masayuki

    2014-08-11

    The wide-gamut system colorimetry has been standardized for ultra-high definition television (UHDTV). The chromaticities of the primaries are designed to lie on the spectral locus to cover major standard system colorimetries and real object colors. Although monochromatic light sources are required for a display to perfectly fulfill the system colorimetry, highly saturated emission colors using recent quantum dot technology may effectively achieve the wide gamut. This paper presents simulation results on the chromaticities of highly saturated non-monochromatic light sources and gamut coverage of real object colors to be considered in designing wide-gamut displays with color filters for the UHDTV.

  20. Analysis of boron dilution in a four-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sun, J.G.; Sha, W.T. [Argonne National Lab., IL (United States)

    1995-03-01

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four-loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant is at hot standby and the reactor coolant system has been heated to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated and a sensitivity study is performed to assess the accuracy of the numerical modeling of the geometry of the reactor coolant system.

  1. Primary antiphospholipid syndrome progressing to systemic lupus erythematosus: a case report

    Directory of Open Access Journals (Sweden)

    Rocco Manganelli

    2013-04-01

    Full Text Available Introduction: Primary antiphospholipid syndrome (APS is a thrombophilic disease that should be suspected in the presence of thrombotic events associated with hematologic abnormalities such as thrombocytopenia and prolongation of the activated partial thromboplastin time. The diagnosis must be confirmed by the demonstration of autoantibodies directed against anionic phospholipids and/or phospholipid-binding proteins. The disease can cause arterial thrombosis in any vascular district, including those of the kidney and central nervous system. Case report: In 2006 a 29-year-old male presented with kidney and brain involvement that was attributed to primary APS. The clinical diagnosis was confirmed by the results of a renal biopsy, which excluded the presence of systemic lupus erythematosus (SLE. The patient remained stable through 32 months of follow-up and then developed a malar rash with deteriorating renal function, decreasing platelet count, and reduced complement levels. Serological studies revealed positivity for ANA (homogeneous pattern, dsDNA, ACA, and beta-2-glycoprotein-1 antibodies. The diagnosis was revised to APS secondary to SLE. Conclusions: A diagnosis of primary APS should not be considered permanent: progression to SLE can occur, in some cases years after the original diagnosis. This case highlights the importance of ongoing follow-up of patients diagnosed with primary APS to detect changes that herald the emergence of SLE.

  2. Primary solidification phases of the Sn-rich Sn-Ag-Cu-Ni quaternary system

    Science.gov (United States)

    Chang, Cheng-An; Chen, Sinn-Wen; Chiu, Chen-Nan; Huang, Yu-Chih

    2005-08-01

    The eutectic and near-eutectic Sn-Ag-Cu solders are the most promising lead-free solders, and nickel is frequently used as the barrier layer material. Nickel dissolves into the molten Sn-Ag-Ni alloy during the soldering process, and the ternary solder becomes a Sn-Ag-Cu-Ni quaternary melt near the nickel substrate. Liquidus projection is the projection of the liquidus trough and it delineates the boundaries of various primary solidification phases. Information of liquidus projection is helpful for understanding the alloys’ solidification behavior. This study prepared the Sn-Ag-Cu-Ni alloys of various compositions at the Sn-rich corner. The alloys were melted at higher temperatures and solidified in air. The solidified alloys were metallographically examined to determine the phases formed, especially the primary solidification phases. No ternary or quaternary compounds were found. The knowledge of the primary solidification phases, phase formation sequences, and reaction temperatures determined in this study were put together with all of the available liquidus projections of the constituent ternary systems to determine the primary solidification phases of the quaternary Sn-Ag-Cu-Ni system at the Sn-rich corner.

  3. Evaluation of primary stability of innovated orthodontic miniscrew system (STS): An ex-vivo study

    Science.gov (United States)

    Seifi, Massoud

    2016-01-01

    Background Stability is determined as one of the requirements in use of Temporary Anchorage Devices (TAD) in orthodontics. Miniscrew has been a widely used Bone Anchor. Compared with mini-implant that necessitates osseointegration; mechanical retention is a determining factor for primary stability of miniscrew. Studies investigated various ways to increase primary stability. The aim of this study is to introduce a new configuration of miniscrew system which is believed to obtain more primary stability. Material and Methods Freshly ovine mandibles were cut in blocks. Twenty-seven miniscrews (diameter 1.6 × 8 mm; G2, Dual Top Anchor System, Jeil Medical, Seoul, Korea) were inserted in the blocks and divided in 2 experimental groups: single miniscrew and the innovated design “Seifi Twin Screw (STS)”. Primary stability was evaluated by Periotest “M”® device. Results Independent t-test showed a significant difference between 2 experimental groups in periotest evaluation (porthodontic anchorage procedures, miniscrews, temporary anchorage device. PMID:27398174

  4. Increased light-use efficiency sustains net primary productivity of shaded coffee plants in agroforestry system.

    Science.gov (United States)

    Charbonnier, Fabien; Roupsard, Olivier; le Maire, Guerric; Guillemot, Joannès; Casanoves, Fernando; Lacointe, André; Vaast, Philippe; Allinne, Clémentine; Audebert, Louise; Cambou, Aurélie; Clément-Vidal, Anne; Defrenet, Elsa; Duursma, Remko A; Jarri, Laura; Jourdan, Christophe; Khac, Emmanuelle; Leandro, Patricia; Medlyn, Belinda E; Saint-André, Laurent; Thaler, Philippe; Van Den Meersche, Karel; Barquero Aguilar, Alejandra; Lehner, Peter; Dreyer, Erwin

    2017-08-01

    In agroforestry systems, shade trees strongly affect the physiology of the undergrown crop. However, a major paradigm is that the reduction in absorbed photosynthetically active radiation is, to a certain extent, compensated by an increase in light-use efficiency, thereby reducing the difference in net primary productivity between shaded and non-shaded plants. Due to the large spatial heterogeneity in agroforestry systems and the lack of appropriate tools, the combined effects of such variables have seldom been analysed, even though they may help understand physiological processes underlying yield dynamics. In this study, we monitored net primary productivity, during two years, on scales ranging from individual coffee plants to the entire plot. Absorbed radiation was mapped with a 3D model (MAESPA). Light-use efficiency and net assimilation rate were derived for each coffee plant individually. We found that although irradiance was reduced by 60% below crowns of shade trees, coffee light-use efficiency increased by 50%, leaving net primary productivity fairly stable across all shade levels. Variability of aboveground net primary productivity of coffee plants was caused primarily by the age of the plants and by intraspecific competition among them (drivers usually overlooked in the agroforestry literature) rather than by the presence of shade trees. © 2017 John Wiley & Sons Ltd.

  5. A clinicopathologic analysis of primary central nervous system lymphomatoid granulomatosis: case report and literature review

    Directory of Open Access Journals (Sweden)

    FU Yong-juan

    2012-10-01

    Full Text Available Objective To investigate the clinical, neuroimaging and histopathological features of primary central nervous system lymphomatoid granulomatosis (LG. Methods The clinical manifestation, neuroimaging, histopathological and biological features of a patient with primary central nervous system LG were presented, and the related literatures were reviewed. Results A 57-year-old male presented with memory impairment, weak in orientation, calculation, apprehension and judgment for 3 months. Magnetic resonance imaging (MRI showed space-occupying lesions in bilateral frontal lobes, with T1WI isointensity and T2WI hyperintensity, and the enhancement was irregular. The lesion was slight expansive with yellow surface and gray-white section in color and soft texture and abundant blood supply. Microscopically, the lesion was characterized by angiocentric and angiodestructive lymphoproliferation, partly showed the structure of LG characterized by T cell predominant proliferation, macrophage infiltration, astrocyte activation, small vessel proliferation and hyalinization, and partly showed the structure of lymphoma characterized by diffuse atypical B cell proliferation, with IgK monoclonal production. Epstein-Barr virus (EBV was negative. Conclusion As a precursor disease of lymphoma, LG should be considered in the differential diagnosis of both diffuse and multifocal lesions of the central nervous system. The relavance between primary central nervous system LG and EBV infection should be further discussed.

  6. The improvement of PWR(OPR-1000) Local Control Pannel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joo-Youl; Kim, Min-Soo; Kim, Kyung-Min; Lee, Jun-Kou [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The malfunction of feature in NPP could be occurred by physical aging, electrical false signal and natural disaster. The first recognition of malfunction is almost done by alarm system. Due to the importance of alarm system, design basis of alarm system is described in FSAR 18.1.4.20(alarm system design review). Operators can recognize malfunction of feature and importance of alarm in short distance. The sound of alarm is also changed depending on frequency so it contributes recognition of alarm. This system is not helpful in recognition of alarm for filed operators. In this study, the way that FSAR(priority of alarm and color indication) is also applied on local control is suggested. The alarm sound considering field situation, alarm name, status indication in circuit breaker are suggested to improve overall local control panel. These can contribute to safety operation. This paper is made from improvement items of local control panel in the sight of field operator. The research of local panel is necessary to apply these improvements and the collaboration of related department is also needed. In this study, The alarm sound considering field situation, alarm name, status indication in circuit breaker are suggested to improve overall local control panel based on Hanul Unit 6. If the improvement is applied, the qualitative effect of safe operation will be increased, and fatigue of work stress will be lower.

  7. Analysis of boron dilution in a four-loop PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sun, J.G.; Sha, W.T.

    1995-12-31

    Thermal mixing and boron dilution in a pressurized water reactor were analyzed with COMMIX codes. The reactor system was the four loop Zion reactor. Two boron dilution scenarios were analyzed. In the first scenario, the plant is in cold shutdown and the reactor coolant system has just been filled after maintenance on the steam generators. To flush the air out of the steam generator tubes, a reactor coolant pump (RCP) is started, with the water in the pump suction line devoid of boron and at the same temperature as the coolant in the system. In the second scenario, the plant is at hot standby and the reactor coolant system has been heated up to operating temperature after a long outage. It is assumed that an RCP is started, with the pump suction line filled with cold unborated water, forcing a slug of diluted coolant down the downcomer and subsequently through the reactor core. The subsequent transient thermal mixing and boron dilution that would occur in the reactor system is simulated for these two scenarios. The reactivity insertion rate and the total reactivity are evaluated.

  8. Primary health care and public health: foundations of universal health systems.

    Science.gov (United States)

    White, Franklin

    2015-01-01

    The aim of this review is to advocate for more integrated and universally accessible health systems, built on a foundation of primary health care and public health. The perspective outlined identified health systems as the frame of reference, clarified terminology and examined complementary perspectives on health. It explored the prospects for universal and integrated health systems from a global perspective, the role of healthy public policy in achieving population health and the value of the social-ecological model in guiding how best to align the components of an integrated health service. The importance of an ethical private sector in partnership with the public sector is recognized. Most health systems around the world, still heavily focused on illness, are doing relatively little to optimize health and minimize illness burdens, especially for vulnerable groups. This failure to improve the underlying conditions for health is compounded by insufficient allocation of resources to address priority needs with equity (universality, accessibility and affordability). Finally, public health and primary health care are the cornerstones of sustainable health systems, and this should be reflected in the health policies and professional education systems of all nations wishing to achieve a health system that is effective, equitable, efficient and affordable.

  9. Emergy assessment of three home courtyard agriculture production systems in Tibet Autonomous Region, China.

    Science.gov (United States)

    Guan, Fa-Chun; Sha, Zhi-Peng; Zhang, Yu-Yang; Wang, Jun-Feng; Wang, Chao

    2016-08-01

    Home courtyard agriculture is an important model of agricultural production on the Tibetan plateau. Because of the sensitive and fragile plateau environment, it needs to have optimal performance characteristics, including high sustainability, low environmental pressure, and high economic benefit. Emergy analysis is a promising tool for evaluation of the environmental-economic performance of these production systems. In this study, emergy analysis was used to evaluate three courtyard agricultural production models: Raising Geese in Corn Fields (RGICF), Conventional Corn Planting (CCP), and Pea-Wheat Rotation (PWR). The results showed that the RGICF model produced greater economic benefits, and had higher sustainability, lower environmental pressure, and higher product safety than the CCP and PWR models. The emergy yield ratio (EYR) and emergy self-support ratio (ESR) of RGICF were 0.66 and 0.11, respectively, lower than those of the CCP production model, and 0.99 and 0.08, respectively, lower than those of the PWR production model. The impact of RGICF (1.45) on the environment was lower than that of CCP (2.26) and PWR (2.46). The emergy sustainable indices (ESIs) of RGICF were 1.07 and 1.02 times higher than those of CCP and PWR, respectively. With regard to the emergy index of product safety (EIPS), RGICF had a higher safety index than those of CCP and PWR. Overall, our results suggest that the RGICF model is advantageous and provides higher environmental benefits than the CCP and PWR systems.

  10. Emergy assessment of three home courtyard agriculture production systems in Tibet Autonomous Region, China*

    Science.gov (United States)

    Guan, Fa-chun; Sha, Zhi-peng; Zhang, Yu-yang; Wang, Jun-feng; Wang, Chao

    2016-01-01

    Home courtyard agriculture is an important model of agricultural production on the Tibetan plateau. Because of the sensitive and fragile plateau environment, it needs to have optimal performance characteristics, including high sustainability, low environmental pressure, and high economic benefit. Emergy analysis is a promising tool for evaluation of the environmental-economic performance of these production systems. In this study, emergy analysis was used to evaluate three courtyard agricultural production models: Raising Geese in Corn Fields (RGICF), Conventional Corn Planting (CCP), and Pea-Wheat Rotation (PWR). The results showed that the RGICF model produced greater economic benefits, and had higher sustainability, lower environmental pressure, and higher product safety than the CCP and PWR models. The emergy yield ratio (EYR) and emergy self-support ratio (ESR) of RGICF were 0.66 and 0.11, respectively, lower than those of the CCP production model, and 0.99 and 0.08, respectively, lower than those of the PWR production model. The impact of RGICF (1.45) on the environment was lower than that of CCP (2.26) and PWR (2.46). The emergy sustainable indices (ESIs) of RGICF were 1.07 and 1.02 times higher than those of CCP and PWR, respectively. With regard to the emergy index of product safety (EIPS), RGICF had a higher safety index than those of CCP and PWR. Overall, our results suggest that the RGICF model is advantageous and provides higher environmental benefits than the CCP and PWR systems. PMID:27487808

  11. A practical guide for the diagnosis of primary enteric nervous system disorders

    DEFF Research Database (Denmark)

    Schäppi, M G; Staiano, A; Milla, P J

    2013-01-01

    OBJECTIVE: Primary gastrointestinal neuropathies are a heterogeneous group of enteric nervous system (ENS) disorders that continue to cause difficulties in diagnosis and histological interpretation. Recently, an international working group published guidelines for histological techniques and repo......OBJECTIVE: Primary gastrointestinal neuropathies are a heterogeneous group of enteric nervous system (ENS) disorders that continue to cause difficulties in diagnosis and histological interpretation. Recently, an international working group published guidelines for histological techniques...... and reporting, along with a classification of gastrointestinal neuromuscular pathology. The aim of this article was to review and summarize the key issues for pediatric gastroenterologists on the diagnostic workup of congenital ENS disorders. In addition, we provide further commentary on the continuing...

  12. Primary angiitis of the central nervous system with diffuse cerebral mass effect and giant cells.

    LENUS (Irish Health Repository)

    Kinsella, J A

    2012-02-01

    Primary angiitis of the central nervous system (PACNS), also called primary CNS vasculitis, is an idiopathic inflammatory condition affecting only intracranial and spinal cord vessels, particularly medium-sized and smaller arteries and arterioles. Angiography and histopathology typically do not reveal evidence of systemic vasculitis.(1,2) Histopathology usually reveals granulomatous inflammation affecting arterioles and small arteries of the parenchyma and\\/or leptomeninges, similar to that seen in Takayasu\\'s or giant cell arteritis.(1-3) We report a patient with biopsy-proven PACNS with giant cells and cerebral mass effect on MRI. Magnetic resonance angiography and cerebral angiography appeared normal and there was no evidence of extracranial vasculitis.

  13. Intravascular Lymphomatosis Mimicking Primary Central Nervous System Lymphoma: A Case Report and Literature Review

    Directory of Open Access Journals (Sweden)

    Masahiro Oomura

    2014-03-01

    Full Text Available We herein report a 75-year-old female patient with intravascular lymphomatosis (IVL who presented with fever of unknown origin. Examination, including contrast-enhanced CT and 67Ga scintigraphy, failed to show any lesions. Her blood levels of lactate dehydrogenase and soluble interleukin-2 receptors were high, suggesting a lymphomatous tumor. A bone marrow puncture was negative, and a random skin biopsy revealed a monoclonal proliferation of naked, large lymphocytes in the vascular space of the subcutaneous tissue, confirming the diagnosis of IVL. MRI, performed 7 weeks after admission, showed a brain mass mimicking primary central nervous system lymphoma. The mass was considered to be a collection of malignant lymphocyte cells invading from the vessels. Without the random skin biopsy, this case may have been misdiagnosed as primary central nervous system lymphoma.

  14. [Legislation on primary care in Brazilian Unified National Health System: document analysis].

    Science.gov (United States)

    Domingos, Carolina Milena; Nunes, Elisabete de Fátima Polo de Almeida; Carvalho, Brígida Gimenez; Mendonça, Fernanda de Freitas

    2016-03-01

    A reflection on Brazil's legislation for primary care helps understand the way health policy is implemented in the country. This study focuses on the legal provisions aimed at strengthening primary care, drawing on an analysis of documents from the Ministry of Health's priority actions, programs, and strategies. A total of 224 provisions were identified, in two groups of documents, so-called instituting provisions and complementary provisions. The former include the principles and guidelines of the Brazilian Unified National Health System (SUS) and also involve the expansion of actions. Financing was a quantitatively central theme, especially in the complementary provisions. The analysis led to reflection on the extent to which these strategies can induce linkage between health system managers and civil society in building a political project resulting in improvements and meeting the population's health needs.

  15. Primary central nervous system peripheral T-cell lymphoma in a child.

    Science.gov (United States)

    Gualco, Gabriela; Wludarski, Sheila; Hayashi-Silva, Luciana; Medeiros Filho, Plinio; Veras, Geni; Bacchi, Carlos Eduardo

    2010-01-01

    A 10-year-old Caucasian boy was admitted to the hospital with a 3-month history of headache, vomiting, ataxia, and right amaurosis. A magnetic resonance imaging (MRI) showed a solid, expansive, parasagittal mass in the right parietal hemisphere that extended sagitally to include the optical chiasm. The lesion was considered unresectable. Histology and immunophenotyping of biopsy tissue revealed characteristics of peripheral T-cell lymphoma. No other anatomical region, including bone marrow, was compromised. Primary T-cell lymphomas of the central nervous system are rare, especially in childhood. Here, we describe the rapidly deteriorating and fatal clinical course of a boy with a primary T-cell lymphoma in the central nervous system.

  16. INTEGRATED DRILLING SYSTEM USING MUD ACTUATED DOWN HOLE HAMMER AS PRIMARY ENGINE

    Energy Technology Data Exchange (ETDEWEB)

    John V. Fernandez; David S. Pixton

    2005-12-01

    A history and project summary of the development of an integrated drilling system using a mud-actuated down-hole hammer as its primary engine are given. The summary includes laboratory test results, including atmospheric tests of component parts and simulated borehole tests of the hammer system. Several remaining technical hurdles are enumerated. A brief explanation of commercialization potential is included. The primary conclusion for this work is that a mud actuated hammer can yield substantial improvements to drilling rate in overbalanced, hard rock formations. A secondary conclusion is that the down-hole mud actuated hammer can serve to provide other useful down-hole functions including generation of high pressure mud jets, generation of seismic and sonic signals, and generation of diagnostic information based on hammer velocity profiles.

  17. Primary sarcomas of the central nervous system: UCSF experience (1985-2005

    Directory of Open Access Journals (Sweden)

    Tarık TİHAN

    2007-01-01

    Full Text Available Sarcomas constitute less than 2% of all cancers, and are a highly diverse group of neoplasms. Primary sarcomas of the central nervous system (CNS are even less common, and our experience is limited by lack of studies with sufficient size that can address the challenges in predicting behavior or management. It is critical to recognize the characteristics of these uncommon neoplasms and to develop better predictors for prognosis and behavior.We have conducted a search of the UCSF Department of Pathology and UCSF Cancer Center Registry databases for all primary CNS sarcomas that were diagnosed and treated between 1985 and 2005. Hemangiopericytomas were included, so were the solitary fibrous tumors due to their close association with the former. We excluded all cases of metastatic sarcomas, chordomas, sarcomatoid variants of all neuroepithelial neoplasms, Ewing’s sarcomas and other embryonal tumors. In addition, we have identified all soft tissue sarcomas diagnosed and treated during the same period. There were 43 primary CNS neoplasms that fulfilled the inclusion criteria. At the same time, we identified 1706 sarcomas primary to the soft tissue. Primary CNS sarcomas included 16 hemangiopericytomas, 15 chondrosarcomas, 3 solitary fibrous tumors, 3 osteosarcomas, 2 leiomyosarcomas, 2 undifferentiated sarcomas, 1 histiocytic sarcoma, and 1 fibrosarcoma. There was a distinctly higher frequency of hemangiopericytoma in the CNS compared to soft tissue. In addition, a group of low grade, parasagittal chondrosarcomas were noted for their highly indolent biological behavior. Unlike some previous series, our cohort was devoid of angiosarcoma and malignant fibrous histiocytoma. This study underscores the limitations of single institutional series, and highlights the value of multi-institutional studies to understand and better treat primary CNS sarcomas.

  18. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Energy Technology Data Exchange (ETDEWEB)

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  19. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  20. Accelerated IGA/SCC testing of Alloy 600 in contaminated PWR environments

    Energy Technology Data Exchange (ETDEWEB)

    Miglin, B.P.; Sarver, J.M. [Babcock & Wilcox R& D Division, Alliance, OH (United States); Aoki, K. [NFI, Osaka (Japan); Koch, D.W. [Babcock & Wilcox Nuclear Services, Lynchburg, VA (United States); Takamatsu, H. [Kansai Electric, Osaka (Japan)

    1992-12-31

    An accelerated corrosion test (360{degrees}C for 2000 hrs) was performed on C-ring specimens machined from one heat of Alloy 600 tubing in the mill-annealed condition. The specimens were exposed to secondary-side pressurized-water-reactor (PWR) solutions contaminated with lead, sulfur, silicon, and a combination of these contaminants. Where possible, MULTEQ calculations were performed to determine the chemical concentrations so that a constant elevated-temperature pH of 4.5 was achieved. This test was designed to examine the ability of these contaminants to cause intergranular attack and/or stress corrosion in stressed Alloy 600 tubing. The results from this test demonstrated that under the test conditions used, lead-contaminated PWR secondary water induces and propagates intergranular attack (IGA) and stress corrosion cracking (SCC) in Alloy 600. Attack was intergranular; the degree of attack did not vary in the liquid or vapor portions of the test environments. Although attack was more severe at higher stresses, significant attack was observed in samples stressed to the typical operating stress. Solutions of only sulfur and only silicon displayed no initiation or propagation of either IGA or SCC. However, the solution containing all three contaminants caused attack with identical morphology to that observed in the lead-contaminated solution.

  1. Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

    Energy Technology Data Exchange (ETDEWEB)

    Cizelj, Leon [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)]. E-mail: leon.cizelj@ijs.si; Koncar, Bostjan [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia); Leskovar, Matjaz [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)

    2006-08-15

    When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures. The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.

  2. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  3. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  4. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  5. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  6. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  7. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  8. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  9. Development of In-pile Plug Assembly and Primary Shutter for Cold Neutron Guide System

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Won; Cho, Yeong Garp; Ryu, Jeong Soo; Lee, Jung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    The HANARO, a 30 MW multi-purpose research reactor in Korea, will be equipped with a neutron guide system, in order to transport cold neutrons from the neutron source to the neutron scattering instruments in the neutron guide hall near the reactor building. The neutron guide system of HANARO consists of the in-pile plug assembly with in-pile guides, the primary shutter with in-shutter guides, the neutron guides in the guide shielding room with dedicated secondary shutters, and the neutron guides connected to the instruments in the neutron guide hall. The functions of the in-pile plug assembly are to shield the reactor environment from a nuclear radiation and to support the neutron guides and maintain them precisely oriented. The primary shutter is a mechanical device to be installed just after the in-pile plug assembly, which stops neutron flux on demand. This report describes the mechanical design, fabrication, and installation procedure of the in-pile plug assembly and the primary shutter for the neutron guide system at HANARO. A special tool and procedure for a replacement of in-pile plug and guide cassette is also presented with the interface condition in the reactor hall.

  10. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  11. Strengthening primary care as main point of entry to the Argentine health system

    Directory of Open Access Journals (Sweden)

    Carlos Alberto Díaz

    2013-09-01

    Full Text Available Background. The Argentine health system is characterized by fragmentation of care, major spending levels and indicators that beg to improve. It is composed of three subsystems: public, private and social security. Despite the efforts of these subsystems, patients do not perceive primary health care as a valid point of entry to health care, preferring to haphazardly choose from a roster of providers. Purpose. To describe and reframe the challenges facing primary health care in order to become the main point of entry to the Argentine health system. Analysis. The magnitude of health inequities have been amply reported, wherein differences in vulnerability and exposure to disease lead to greater social stratification. This situation makes it necessary to organize a coordinated, effective and sustainable strategy to tackle existing health problems, new challenges and improve equity. Conclusion. Primary health care should become the main point of entry to the Argentine health system. This is a long term endeavor that requires joint and coordinated decisions aimed at strengthening a model of care based on prevention and early detection of disease, together with efficient spending and an improvement in the quality of life of the population.

  12. Primary user localisation and uplink resource allocation in orthogonal frequency division multiple access cognitive radio systems

    KAUST Repository

    Nam, Haewoon

    2015-05-21

    In cognitive radio networks, secondary users (SUs) can share spectrum with primary users (PUs) under the condition that no interference is caused to the PUs. To evaluate the interference imposed to the PUs, the cognitive systems discussed in the literature usually assume that the channel state information (CSI) of the link from a secondary transmitter to a primary receiver (interference link) is known at the secondary transmitter. However, this assumption may often be impractical in cognitive radio systems, since the PUs need to be oblivious to the presence of the SUs. The authors first discuss PU localisation and then introduce an uplink resource allocation algorithm for orthogonal frequency division multiple access-based cognitive radio systems, where relative location information between primary and SUs is used instead of CSI of the interference link to estimate the interference. Numerical and simulation results show that it is indeed effective to use location information as a part of resource allocation and thus a near-optimal capacity is achieved. © The Institution of Engineering and Technology 2015.

  13. Deployment and Validation of a Smart System for Screening of Language Disorders in Primary Care

    Directory of Open Access Journals (Sweden)

    Iván Pau de la Cruz

    2013-06-01

    Full Text Available Neuro-evolutive development from birth until the age of six years is a decisive factor in a child’s quality of life. Early detection of development disorders in early childhood can facilitate necessary diagnosis and/or treatment. Primary-care pediatricians play a key role in its detection as they can undertake the preventive and therapeutic actions requested to promote a child’s optimal development. However, the lack of time and little specific knowledge at primary-care avoid to applying continuous early-detection anomalies procedures. This research paper focuses on the deployment and evaluation of a smart system that enhances the screening of language disorders in primary care. Pediatricians get support to proceed with early referral of language disorders. The proposed model provides them with a decision-support tool for referral actions to trigger essential diagnostic and/or therapeutic actions for a comprehensive individual development. The research was conducted by starting from a sample of 60 cases of children with language disorders. Validation was carried out through two complementary steps: first, by including a team of seven experts from the fields of neonatology, pediatrics, neurology and language therapy, and, second, through the evaluation of 21 more previously diagnosed cases. The results obtained show that therapist positively accepted the system proposal in 18 cases (86% and suggested system redesign for single referral to a speech therapist in three remaining cases.

  14. Breast cancer brain metastases responding to lapatinib plus capecitabine as second-line primary systemic therapy.

    Science.gov (United States)

    Bergen, Elisabeth S; Berghoff, Anna S; Rudas, Margaretha; Preusser, Matthias; Bartsch, Rupert

    2015-06-01

    Brain metastases (BM) are diagnosed in up to 40% of HER2-positive breast cancer patients. Standard treatment includes local approaches such as whole-brain radiotherapy (WBRT), radiosurgery, and neurosurgery. The landscape trial established primary systemic therapy as an effective and safe alternative to WBRT in selected patients with Her2-positive BM. We aim to further focus on the role of systemic therapy in oligosymptomatic patients by presenting this case report. We report on a 50-year-old patient diagnosed with multiple BM 5 years after early breast cancer diagnosis. As the patient was asymptomatic and had a favorable diagnosis-specific GPA score, she received primary systemic treatment with T-DM1. She achieved partial remission within the brain for eight treatment cycles and then progressed despite stable extracranial disease. As the patient remained asymptomatic and refused WBRT, we decided upon trastuzumab, lapatinib plus capecitabine as second-line therapy. Another partial remission of BM was observed; to date, she has received 11 treatment cycles without any sign of disease progression. In this case, WBRT was delayed by at least 14 months, again indicating the activity of systemic treatment in BM. Apparently, in selected patients, BM can be controlled with multiple lines of systemic therapy similar to extracranial disease. Further investigation of systemic treatment approaches is therefore warranted.

  15. Claim criteria of significant events implying the safety of PWR type reactors; Criteres de declaration des evenements significatifs impliquant la surete pour les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-10-15

    There are ten criteria for the declaration of the significant events implying the safety for PWR type reactors. First criterion: Automatic stop of the reactor: manual or automatic, inconvenient starting or not, the function of automatic stop of the reactor, whatever is the state of the reactor, with the exception of the deliberate starting resulting from planned actions. Second criterion: Starting of one of the systems of protection, manual or automatic, inconvenient starting or not, of one of the systems of protection, with the exception of the deliberate starting resulting from planned actions. Third criterion: Disregard of the technical specifications of exploitation (S.T.E ), or an event which would have been able to lead to a disregard of the S.T.E., if the same event had occurred, the installation having been in a different state, any disregard of one or several permanent conditions defined in S.T.E., any disregard of the conditions of a dispensation in S.T.E., any overtaking of periods when it is not prescribed by state of fold, any unavailability provoked outside the conditions planned by the main rules of exploitation, not identified beforehand or identified but untreated according to the prescriptions of the S.T.E. fourth criterion: Internal or external aggression, happening of a natural external phenomenon or in relation with a human activity, or happening of an internal flooding, a fire or another phenomenon susceptible to affect the availability of the equipment important for the safety. Fifth criterion: Act or attempt of act of hostility susceptible to affect the safety of the installation. Sixth criterion: Passage in state of fold in application of the technical specifications of exploitation or the accidental procedures of driving following an unforeseen behavior of the installation. Seventh criterion: Event having cause or being able to cause multiple failures, unavailability of equipment due to the same failure either affecting all the ways of a

  16. Use of email in communication between the Finnish primary healthcare system and general practitioners.

    Science.gov (United States)

    Karhula, Tuula; Kauppila, Timo; Elonheimo, Outi; Brommels, Mats

    2011-01-01

    The volume of emails is rising rapidly everywhere. However, there is no data available concerning how primary healthcare physicians feel about the use of email communication between themselves, with their managers and with other people contacting them. The objective of this study was to find out what the attitudes of primary care physicians are towards email at work. The use of email was studied among a convenience sample of primary healthcare physicians. Physicians thought that email was a good instrument for delivering information but not as an instrument for leadership. Physicians in lead positions thought more often than ordinary general practitioners (GPs) that email is good for information. The leaders used email more actively than other GPs. The contents of the emails received by the GPs differed depending on the site of work. The total number of emails was higher in urban areas than in rural areas. Emails relating to administration, educational information and meeting materials were more often sent in rural than in urban primary healthcare settings. Information about daily work arrangements and about social events were more frequently emailed in urban than in rural surroundings. Email was considered important for information inside the system but a somewhat difficult tool for discussing complicated subjects. Generally, it was agreed that there was some unimportant information filtering through this medium to the target GPs. GPs were uncertain whether important data reached everybody who needed it or not. Still, almost everybody used the email system regularly and the use of it was considered relatively easy. GPs were generally prone to adopt advice and instructions given via email and implemented those in their working routines. The use of the email system was related to technical ability to use the system. The easier the GP thought that the email system was the more he used it. Rural GPs were more critical in applying advice shared via email than their

  17. Cooperative Research and Development of Primary Surface Recuperator for Advanced Microturbine Systems

    Energy Technology Data Exchange (ETDEWEB)

    Escola, George

    2007-01-17

    Recuperators have been identified as key components of advanced gas turbines systems that achieve a measure of improvement in operating efficiency and lead the field in achieving very low emissions. Every gas turbine manufacturer that is studying, developing, or commercializing advanced recuperated gas turbine cycles requests that recuperators operate at higher temperature without a reduction in design life and must cost less. The Solar Cooperative Research and Development of Primary Surface Recuperator for Advanced Microturbine Systems Program is directed towards meeting the future requirements of advanced gas turbine systems by the following: (1) The development of advanced alloys that will allow recuperator inlet exhaust gas temperatures to increase without significant cost increase. (2) Further characterization of the creep and oxidation (dry and humid air) properties of nickel alloy foils (less than 0.13 mm thick) to allow the economical use of these materials. (3) Increasing the use of advanced robotic systems and advanced in-process statistical measurement systems.

  18. Primary frequency regulation with Li-ion battery energy storage system: A case study for Denmark

    DEFF Research Database (Denmark)

    Swierczynski, Maciej Jozef; Stroe, Daniel Ioan; Stan, Ana-Irina

    2013-01-01

    and improving the predictability of the intermittent renewables but also of providing the ancillary services in the future energy markets. However, this is currently difficult to achieve due to high prices of the energy storage systems and difficulties with accurate prediction of the energy storage systems......Meeting ambitious goals of transition to distributed and environmentally-friendly renewable energy generation can be difficult to achieve without energy storage systems due to technical and economical challenges. Moreover, energy storage systems have a high potential of not only smoothing...... on the results obtained from accelerated lifetime testing. The developed Li­-ion battery lifetime model is later a base for the analyses of the economic profitability of the investment in the Li-ion battery energy storage system (BESS), which delivers the primary frequency regulation service on the Danish...

  19. Stress corrosion cracking in the vessel closure head penetrations of French PWR`s; Fissuration par corrosion sous contrainte de penetrations de couvercle de cuve de reacteur nucleaire francais a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Buisine, D.; Cattant, F.; Champredonde, J.; Pichon, C.; Benhamou, C.; Gelpi, A.; Vaindirlis, M.

    1994-01-01

    During a hydrotest in September 1991, part of the statutory decennial in-service inspection, a leak was detected on the vessel head of Bugey 3, which is one of the first 900 MW 3-loop PWR`s in France. This leak was due to a cracked penetration used for a control rod drive mechanism. The investigations performed identified Primary Stress Corrosion Cracking of Alloy 600 as being the origin of this degradation. So a lot of the same design PWR`s are a concern due to this generic problem. In this case, PWSCC was linked to: - hot temperature of the vessel head; - high residual stresses due to the welding process between peripherical penetrations and the vessel head; - sensitivity of forged Alloy 600 used for penetration manufacturing. This following paper will present the cracked analysis based, in particular, on the main results obtained in France on each of these items. These results come from the operating experience, the destructive examinations and the programs which are running on stress analysis and metallurgical characterizations. (authors). 9 figs., 2 tabs.

  20. Clinical analysis of multiple primary malignancies in the digestive system: A hospital-based study

    Institute of Scientific and Technical Information of China (English)

    Hui-Yun Cheng; Cheng-Hsin Chu; Wen-Hsiung Chang; Tzu-Chi Hsu; Shee-Chan Lin; Chuan-Chuan Liu; An-Ming Yang; Shou-Chuan Shih

    2005-01-01

    AIM: To analyze the characteristics of multiple primary malignancies (MPMs) of digestive system; including incidence, types of tumor combinations, time intervals between development of multiple tumors, clinical course,and prognostic factors affecting survival and mortality.METHODS: Data from a total of 129 patients treated from January 1991 to December 2000 for pathologically proved MPMs, including at least one originating from the digestive system, were reviewed retrospectively.RESULTS: Among 129 patients, 120 (93.02%) had two primary cancers and 9 (6.98%) had three primary cancers. The major sites of MPMs of the digestive system were large intestine, stomach, and liver. Associated nondigestive cancers included 40 cases of gynecological cancers, of which 31 were carcinoma of cervix and 10cases of genitourinary cancers, of which 5 were bladder cancers. Other cancers originated from the lung, breast,nasopharynx, larynx, thyroid, brain, muscle, and skin.Reproductive tract cancers, especially cervical, ovarian,bladder, and prostate cancers were the most commonly associated non-GI cancers, followed by cancer of the lung and breasts. Forty-three cases were synchronous, while the rest (86 cases) were metachronous cancers. Staging of MPMs and treatment regimes correlated with the prognosis between survival and non-survival groups.CONCLUSION: As advances in cancer therapy bring about a progressively larger percentage of long-term survivors, the proportion of patients with subsequent primary lesions will increase. Early diagnosis of these lesions, based on an awareness of the possibility of second and third cancers, and multidiscipiinary treatment strategies will substantially increase the survival of these patients.

  1. Component failures that lead to reactor scrams. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  2. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in BWR and PWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Haenninen, H.; Aho-Mantila, I.

    1981-05-01

    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, has been examined both in O2-enriched BWR-conditions (8 ppm O2) and in typical PWR-conditions.

  3. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    Energy Technology Data Exchange (ETDEWEB)

    David Andrs; Ray Berry; Derek Gaston; Richard Martineau; John Peterson; Hongbin Zhang; Haihua Zhao; Ling Zou

    2012-05-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  4. DUPIC fuel fabrication using spent PWR fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2000-12-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  5. DUPIC fuel fabrication using spent PWR fuel at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2001-09-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter 3/4y, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  6. A liquid-metal filling system for pumped primary loop space reactors

    Science.gov (United States)

    Crandall, D. L.; Reed, W. C.

    Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

  7. 压水堆冷凝回流特性的研究%AN INVESTIGATION ON PWR REFLUX CONDENSATION CHARACTERISTICS

    Institute of Scientific and Technical Information of China (English)

    陈听宽; 杨鲁伟

    2000-01-01

    In the high pressure steam-water two-phase flow test system of Xi'an Jiaotong University, the reflux condensation characteristics at PWR SBLOCA are simulated successfully. In experiments, the heat transfer, flow andnoncondensable gas effect are determined. From the tests, the reflux condensation heat transfer is very effective to remove the reactor core decay heat under small temperature difference. Generally, reflux condensation's flow resistance is small and flow is stable. But when countercurrent flow limit is reached, the flow will change to be instable. Noncondensable gas can degrade the heat transfer ability in steam generator, but system can increase the pressure in reactor automatically to remove the decay heat when only a little noncondensable gas exists.%利用高压汽水两相流试验系统模拟压水堆小破口失水事故中冷凝回流传热模式,进行了传热、流动及不凝结气影响的试验。实验表明:冷凝回流传热是一种十分有效的传热模式,它在很小的一、二次侧温差时就能排放大量堆芯余热。冷凝回流系统在正常情况下流动阻力很小且稳定,但在达到回流流动极限后出现不稳定。不凝结气的存在将大大降低蒸汽发生器的传热能力,但一般情况下,系统能自动增加一次侧压力而达到排除余热的目的。

  8. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  9. End effects on elbows subjected to moment loadings. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodabaugh, E.C.; Moore, S.E.

    1982-01-01

    So-called end effects for moment loadings on short-radius and long-radius butt welding elbows of various arc lengths are investigated with a view toward providing more accurate design formulas for critical piping systems. Data developed in this study, along with published information, were used to develop relatively simple design equations for elbows attached at both ends to long sections of straight pipe. These formulas are the basis for an alternate ASME Code procedure for evaluating the bending moment stresses in Class 1 nuclear piping (ASME Code Case N-319). The more complicated problems of elbows with other end conditions, e.g., flanges at one or both ends, are also considered. Comparisons of recently published experimental and theoretical studies with current industrial code design rules for these situations indicate that these rules also need to be improved.

  10. Analysis of the Primary Constraint Conditions of an Efficient Photovoltaic-Thermoelectric Hybrid System

    Directory of Open Access Journals (Sweden)

    Guiqiang Li

    2016-12-01

    Full Text Available Electrical efficiency can be increased by combining photovoltaic (PV and the thermoelectric (TE systems. However, a simple and cursory combination is unsuitable because the negative impact of temperature on PV may be greater than its positive impact on TE. This study analyzed the primary constraint conditions based on the hybrid system model consisting of a PV and a TE generator (TEG, which includes TE material with temperature-dependent properties. The influences of the geometric size, solar irradiation and cold side temperature on the hybrid system performance is discussed based on the simulation. Furthermore, the effective range of parameters is demonstrated using the image area method, and the change trend of the area with different parameters illustrates the constraint conditions of an efficient PV-TE hybrid system. These results provide a benchmark for efficient PV-TEG design.

  11. SALT segmented primary mirror: commissioning capacitive edge sensing system and performance comparison with inductive sensor

    Science.gov (United States)

    Buous, Sebastien; Menzies, John; Gajjar, Hitesh

    2008-07-01

    The SAMS (Segment Alignment Measurement System) is a capacitance-based edge sensing solution for the active alignment of the 10m SALT segmented primary mirror. Commissioning and calibrating the system has been an ongoing task in an attempt to counteract the unfavourable response of the sensors to high humidity conditions and high dust levels. Several solutions were implemented and tested including real-time feedback systems and the application of corrective functions. In parallel with the continuing efforts to improve the performance of the capacitive sensors, we have also been testing a prototype inductive sensor developed by Fogale Nanotech that is of a very similar flexible plate construction. In this paper we present the results obtained and performance gains achieved thus far with the capacitive edge-sensing system as well as a performance comparison of the Fogale inductive sensor to the capacitive edge sensor.

  12. Design and testing of the U.S. Space Station Freedom primary propulsion system

    Science.gov (United States)

    Morano, Joseph S.; Delventhal, Rex A.; Chilcot, Kimberly J.

    1992-07-01

    The primary propulsion system (PPS) for the Space Station Freedom is discussed in terms of salient design characteristics and key testing procedures. The rocket engine modules contain reboost and attitude control thrusters, and their designs are illustrated showing the mounting structures, thruster solenoid valves, and thrust chambers. The propellant tank assembly for storing gaseous N pressurant and hydrazine propellant is described as are the system avionics, thruster solenoid valves, and latching isolation valves. PPS testing conducted on the development systems includes the use of a propulsion-module development unit, a development test article, and system qualification testing. Specific test articles include functional heaters, mass/thermal simulated components, flight-quality structures, and software control operations.

  13. Operation of Grid -Connected Lithium-Ion Battery Energy Storage System for Primary Frequency Regulation

    DEFF Research Database (Denmark)

    Stroe, Daniel Loan; Knap, Vaclav; Swierczynski, Maciej Jozef

    2017-01-01

    Because of their characteristics, which have been continuously improved during the last years, Lithium ion batteries were proposed as an alternative viable solution to present fast-reacting conventional generating units to deliver the primary frequency regulation service. However, even though...... there are worldwide demonstration projects where energy storage systems based on Lithium-ion batteries are evaluated for such applications, the field experience is still very limited. In consequence, at present there are no very clear requirements on how the Lithium-ion battery energy storage systems should...... be operated while providing frequency regulation service and how the system has to re-establish its SOC once the frequency event has passed. Therefore, this paper aims to investigate the effect on the lifetime of the Lithium-ion batteries energy storage system of various strategies for re...

  14. Suggested Operation Grid-Connected Lithium-Ion Battery Energy Storage System for Primary Frequency Regulation

    DEFF Research Database (Denmark)

    Stroe, Daniel Ioan; Knap, Vaclav; Swierczynski, Maciej Jozef;

    2015-01-01

    there are worldwide demonstration projects where energy storage systems based on Lithium-ion batteries are evaluated for such applications, the field experience is still very limited. In consequence, at present there are no very clear requirements on how the Lithium-ion battery energy storage systems should...... be operated while providing frequency regulation service and how the system has to re-establish its SOC once the frequency event has passed. Therefore, this paper aims to investigate the effect on the lifetime of the Lithium-ion batteries energy storage system of various strategies for re......Because of their characteristics, which have been continuously improved during the last years, Lithium ion batteries were proposed as an alternative viable solution to present fast-reacting conventional generating units to deliver the primary frequency regulation service. However, even though...

  15. The panels for primary and secondary mirror reflectors and the Active Surface System for the new Sardinia Radio Telescope

    Science.gov (United States)

    Zacchiroli, G.; Fiocchi, F.; Maccaferri, G.; Morsiani, M.; Orfei, A.; Pernechele, C.; Pisanu, T.; Roda, J.; Vargiu, G.

    In this paper we will describe the panels for the primary and secondary mirror reflectors and the active surface system that will be provided on the Sardinia Radio Telescope. The panels for the primary and secondary mirror have been designed to allow an operating frequency up to 100 GHz. The active surface system will be used to overcome the effect of gravity deformation on the antenna gain and to re-shape the primary mirror in a parabolic form, in order to avoid large phase error contribution on the gain for the highest frequencies placed in the primary focus.

  16. Challenge of Primary Voltage Control in Large Scale Wind Integrated Power System: A Danish Power System Case Study

    DEFF Research Database (Denmark)

    Rather, Zakir Hussain; Chen, Zhe; Thøgersen, Paul

    2013-01-01

    Grid integration of Renewable Energy (RE) at large scale poses vast majority of challenges to secure and stable operation of Power System. This paper presents the challenge of short circuit power and primary voltage control of wind integrated power system where majority of conventional generators...... are replaced by wind generators. The impact of large scale wind integration on fast reactive power support is studied in this paper. Considering both technical and economic aspects, alternatives to address the challenge of dynamic voltage support have also been demonstrated in this paper. A case study...

  17. Environmental standards for primary and secondary containment systems and transfer stations

    Energy Technology Data Exchange (ETDEWEB)

    Maguire, D.M.

    1995-04-01

    Environmental Standards for Primary and Secondary Containment Systems and Transfer Stations will supersede all previous requirements for design of dikes, storage tanks, and transfer stations in order to maintain consistency throughout the Y-12 Plant. This document is organized into six distinct sections, each with a specific purpose. Section I outlines the objectives of the document along with its applications and limitations; this section should be of interest to all readers for essential background information. Section II lists all definitions and is consistent with definitions outlined by environmental regulations. Section III discusses primary containment standards. Section IV outlines secondary containment standards; this section contains the actual standards for the diking of storage tanks and storage containers. Section V discusses transfer station standards. Section VI of this document outlines how exemptions may be granted for specific cases.

  18. Primary biodegradation of veterinary antibiotics in aerobic and anaerobic surface water simulation systems

    DEFF Research Database (Denmark)

    Ingerslev, Flemming; Toräng, Lars; Loke, M.-L.

    2001-01-01

    of the study was to provide rate data for primary biodegradation in the scenario where antibiotics pollute surface waters as a result of run-off from arable land. The source of antibiotics may be application of manure as fertilizer or excreta of grazing animals. Assuming first-order degradation kinetics......The primary aerobic and anaerobic biodegradability at intermediate concentrations (50-5000 mug/l) of the antibiotics olaquindox (OLA), metronidazole (MET), tylosin (TYL) and oxytetracycline (OTC) was studied in a simple shake flask system simulating the conditions in surface waters. The purpose......, ranges of half-lifes for aerobic degradation of the four antibiotics studied were 4-8 days (OLA), 9.5-40 days (TYL), 14-104 days (MET) and 42-46 days (OTC). OLA and OTC were degraded with no initial lag phase whereas lag phases from 2 to 34 days (MET) and 31 to 40 days (TYL) were observed for other...

  19. Planning guidance for nuclear-power-plant decontamination. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Munson, L.F.; Divine, J.R.; Martin, J.B.

    1983-06-01

    Direct and indirect costs of decontamination are considered in the benefit-cost analysis. A generic form of the benefit-cost ratio is evaluated in monetary and nonmonetary terms, and values of dollar per man-rem are cited. Federal and state agencies that may have jurisiction over various aspects of decontamination and waste disposal activities are identified. Methods of decontamination, their general effectiveness, and the advantages and disadvantages of each are outlined. Dilute or concentrated chemical solutions are usually used in-situ to dissolve the contamination layer and a thin layer of the underlying substrate. Electrochemical techniques are generally limited to components but show high decontamination effectiveness with uniform corrosion. Mechanical agents are particularly appropriate for certain out-of-system surfaces and disassembled parts. These processes are catagorized and specific concerns are discussed. The treatment, storage, and disposal or discharge or discharge of liquid, gaseous, and solid wastes generated during the decontamination process are discussed. Radioactive and other hazardous chemical wastes are considered. The monitoring, treatment, and control of radioactive and nonradioactive effluents, from both routine operations and possible accidents, are discussed. Protecting the health and safety of personnel onsite during decontamination is of prime importance and should be considered in each facet of the decontamination process. The radiation protection philosophy of reducing exposure to levels as low as reasonably achievable should be stressed. These issues are discussed.

  20. Development of computer assisted instruction for PWR technology

    Energy Technology Data Exchange (ETDEWEB)

    Chihara, Yasutaka [Kansai Electric Power Co., Inc., Osaka (Japan); Kubo, Setsuo; Ninomiya, Toshiaki

    2001-02-01

    Because of successive formation of large accidents such as critical accident in JCO, exfoliation of concrete in JR, some opinions were increased on anxiety of decline on technical capacity in Japan. Required for a long term from its construction to its operation, a nuclear power plant is now at a state difficult to practise its technical tradition at its working field. In order to hand down technique with good quality, it is essential to receive a wrong experience such as trouble as a noble inheritance, to exert without losing it by its forgetting. The Wakasa Branch of the Kansai Electric Power Co., Ltd. has exerted renewal of its instruments and so on laid stress on their reliability and safety by activating its nuclear technical capacity and some instructions on past trouble. The Branch edited these informations and experiences to a data of CD ROM aiming at smooth technical tradition for next generation. Process on transition of the nuclear technical capacity is laid from beginning of nuclear power plant to APWR. By using the data, four technical instructions on volume of the central control board, volume of reactor vessel (R/V), volume of steam generator (S/G), and volume of RCP have been already made-up. Here were introduced on volumes of R/V and S/G belonging to their mechanical systems. (G.K.)

  1. Neural systems underlying aversive conditioning in humans with primary and secondary reinforcers

    Directory of Open Access Journals (Sweden)

    Mauricio R Delgado

    2011-05-01

    Full Text Available Money is a secondary reinforcer commonly used across a range of disciplines in experimental paradigms investigating reward learning and decision-making. The effectiveness of monetary reinforcers during aversive learning and its neural basis, however, remains a topic of debate. Specifically, it is unclear if the initial acquisition of aversive representations of monetary losses depends on similar neural systems as more traditional aversive conditioning that involves primary reinforcers. This study contrasts the efficacy of a biologically defined primary reinforcer (shock and a socially defined secondary reinforcer (money during aversive learning and its associated neural circuitry. During a two-part experiment, participants first played a gambling game where wins and losses were based on performance to gain an experimental bank. Participants were then exposed to two separate aversive conditioning sessions. In one session, a primary reinforcer (mild shock served as an unconditioned stimulus (US and was paired with one of two colored squares, the conditioned stimuli (CS+ and CS-, respectively. In another session, a secondary reinforcer (loss of money served as the US and was paired with one of two different CS. Skin conductance responses were greater for CS+ compared to CS- trials irrespective of type of reinforcer. Neuroimaging results revealed that the striatum, a region typically linked with reward-related processing, was found to be involved in the acquisition of aversive conditioned response irrespective of reinforcer type. In contrast, the amygdala was involved during aversive conditioning with primary reinforcers, as suggested by both an exploratory fMRI analysis and a follow-up case study with a patient with bilateral amygdala damage. Taken together, these results suggest that learning about potential monetary losses may depend on reinforcement learning related systems, rather than on typical structures involved in more biologically based

  2. Quality of documentation of electronic medical information systems at primary health care units in Alexandria, Egypt.

    Science.gov (United States)

    Noureldin, M; Mosallam, R; Hassan, S Z

    2014-03-13

    Limited data are available about the implementation of electronic records systems in primary care in developing countries. The present study aimed to assess the quality of documentation in the electronic medical records at primary health care units in Alexandria, Egypt and to elicit physician's feedback on barriers and facilitators to the system. Data were collected at 7 units selected randomly from each administrative region and in each unit 50 paper-based records and their corresponding e-records were randomly selected for patients who visited the unit in the first 3 months of 2011. Administrative data were almost complete in both paper and e-records, but the completeness of clinical data varied between 60.0% and 100.0% across different units and types of record. The accuracy rate of the main diagnosis in e-records compared with paper-based records ranged between 44.0% and 82.0%. High workload and system complexity were the most frequently mentioned barriers to implementation of the e-records system.

  3. Cardiac involvement in primary systemic vasculitis and potential drug therapies to reduce cardiovascular risk.

    Science.gov (United States)

    Misra, Durga Prasanna; Shenoy, Sajjan N

    2017-01-01

    Cardiac involvement is common in primary systemic vasculitides and may be due to direct effect of the disease on the heart or due to therapy. We shall review involvement of the heart in the various forms of primary systemic vasculitis. Among anti-neutrophil cytoplasmic antibody-associated vasculitis (AAV), eosinophilic granulomatosis with polyangiitis most commonly involves the heart. Involvement of the heart confers poorer prognosis in AAV, which is also complicated by increased risk of cardiovascular events. Kawasaki's disease (KD) is the most common form of medium-vessel vasculitis to affect the heart, with coronary artery aneurysms being the most common manifestation. These predispose patients with KD to develop premature ischemic heart disease. Takayasu's arteritis is the most common large-vessel vasculitis to involve the heart and can result in aortic incompetence, myocarditis, or coronary heart disease. Involvement of the heart in Behcet's disease is usually in the form of intracardiac mass lesions, thrombosis, or endomyocardial fibrosis. Drugs used in the treatment of systemic vasculitis influence the risk of developing cardiovascular events. Corticosteroid therapy has been shown to increase the risk of myocardial infarction, whereas methotrexate, azathioprine, mycophenolate mofetil, rituximab, and anti-tumor necrosis alpha agents favorably modulate the risk of cardiovascular events, predominantly by dampening systemic inflammation. Awareness of cardiac involvement in vasculitis and accelerated cardiovascular risk in these patients should help clinicians to maximize the modulation of modifiable risk factors for heart disease in these individuals.

  4. EHR systems in the Spanish Public Health National System: the lack of interoperability between primary and specialty care.

    Science.gov (United States)

    de la Torre-Díez, Isabel; González, Sandra; López-Coronado, Miguel

    2013-02-01

    One of the problems of the Spanish Public Health National System is the lack of interoperability in the implemented Electronic Health Records (EHRs) systems in primary and specialty care. There is a deficiency in the electronic health systems that store the data of primary care patients, so one of the basic problems that prevent that every hospital and health center working on the same method is that deficiency. In this paper we research on this problem and to give expression to a series of solutions to it. Bibliographic material in this work has been obtained mainly from MEDLINE source. Additionally, due to the lack of information and privacy about the different EHRs systems, we have resorted to making direct contact with the organizations that have implemented those systems and technological providers. Two solutions have been propounded given several aspects for a feasibility study. The first solution is based upon in the execution of backups in different EHRs databases, which implies a huge economical and infrastructure development. The second of these solutions so that due to the creation of protocols by means of Cloud Computing Technologies. It is crucial the need to reach a homogeneity concerning to the storage of patients clinical data. On the results achieved we can emphasize that maybe the main problems are not the economical handicaps or the large technological development needed, but, as for Health each Region manages its own competences, each one governs with independent policies and decisions.

  5. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

  6. [Perceptions of primary care physicians in Madrid on the austerity measures in the health care system].

    Science.gov (United States)

    Heras-Mosteiro, Julio; Otero-García, Laura; Sanz-Barbero, Belén; Aranaz-Andrés, Jesús María

    2016-01-01

    To address the current economic crisis, governments have promoted austerity measures that have affected the taxpayer-funded health system. We report the findings of a study exploring the perceptions of primary care physicians in Madrid (Spain) on measures implemented in the Spanish health system. We carried out a qualitative study in two primary health care centres located in two neighbourhoods with unemployment and migrant population rates above the average of those in Madrid. Interviews were conducted with 12 primary health care physicians. Interview data were analysed by using thematic analysis and by adopting some elements of the grounded theory approach. Two categories were identified: evaluation of austerity measures and evaluation of decision-making in this process. Respondents believed there was a need to promote measures to improve the taxpayer-funded health system, but expressed their disagreement with the measures implemented. They considered that the measures were not evidence-based and responded to the need to decrease public health care expenditure in the short term. Respondents believed that they had not been properly informed about the measures and that there was adequate professional participation in the prioritization, selection and implementation of measures. They considered physician participation to be essential in the decision-making process because physicians have a more patient-centred view and have first-hand knowledge of areas requiring improvement in the system. It is essential that public authorities actively involve health care professionals in decision-making processes to ensure the implementation of evidence-based measures with strong professional support, thus maintaining the quality of care. Copyright © 2016 SESPAS. Published by Elsevier Espana. All rights reserved.

  7. A framework to support team-based models of primary care within the Australian health care system.

    Science.gov (United States)

    Naccarella, Lucio; Greenstock, Louise N; Brooks, Peter M

    2013-09-01

    Health systems with strong primary care orientations are known to be associated with improved equity, better access for patients to appropriate services at lower costs, and improved population health. Team-based models of primary care have emerged in response to health system challenges due to complex patient profiles, patient expectations and health system demands. Successful team-based models of primary care require a combination of interprofessional education and learning; organisational and management policies and systems; and practice support systems. To ensure evidence is put into practice, we propose a framework comprising five domains (theory, implementation, infrastructure, sustainability and evaluation) to assist policymakers, educators, researchers, managers and health professionals in supporting team-based models of primary care within the Australian health care system.

  8. EVALUATION OF BEST AVAILABLE CONTROL TECHNOLOGY FOR TOXICS -TBACT- DOUBLE SHELL TANK FARMS PRIMARY VENTILATION SYSTEMS SUPPORTING WASTE TRANSFER OPERATIONS

    Energy Technology Data Exchange (ETDEWEB)

    HAAS CC; KOVACH JL; KELLY SE; TURNER DA

    2010-06-24

    This report is an evaluation of Best Available Control Technology for Toxics (tBACT) for installation and operation of the Hanford double shell (DST) tank primary ventilation systems. The DST primary ventilation systems are being modified to support Hanford's waste retrieval, mixing, and delivery of single shell tank (SST) and DST waste through the DST storage system to the Waste Treatment and Immobilizaiton Plant (WTP).

  9. EVALUATION OF BEST AVAILABLE CONTROL TECHNOLOGY FOR TOXICS (TBACT) DOUBLE SHELL TANK FARMS PRIMARY VENTILATION SYSTEM SUPPORTING WASTE TRANSFER OPERATIONS

    Energy Technology Data Exchange (ETDEWEB)

    KELLY SE; HAASS CC; KOVACH JL; TURNER DA

    2010-06-03

    This report is an evaluation of Best Available Control Technology for Toxics (tBACT) for installation and operation of the Hanford double shell (DST) tank primary ventilation systems. The DST primary ventilation systems are being modified to support Hanford's waste retrieval, mixing, and delivery of single shell tank (SST) and DST waste throught the DST storage system to the Waste Treatment and Immobilization Plant (WTP).

  10. A study of solute transport of radiolysis products in crud and its effects on crud growth on PWR fuel pin

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Justin H. [BNF Consulting (United States); Kim, Seung Jun, E-mail: skim@lanl.gov [Mechanical and Thermal Engineering Group (AET-1), Los Alamos National Laboratory (United States); Jones, Barclay G. [Department of Nuclear Plasma Radiological Engineering, University of Illinois Urbana-Champaign (United States)

    2016-04-15

    Highlights: • We model a 3-D numerical solute transport within crud deposit on PWR fuel pin. • Source term effect from radiolysis yield and recombination is minimal. • Lower crud porosity leads substantially higher concentration of solutes. • Thicker crud deposit generates substantially higher concentration of solutes. • High concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) can be directly related to corrosion issues on fuel cladding. - Abstract: This research examines the concentration of radiolysis species (H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2}) over the porous crud layer using a three dimensional time dependent solute transport model. A Monte Carlo random walk technique is adopted to simulate the transport behavior of the different species with various parametric studies of source term, crud thickness, and crud porosity. Particularly, this model employs a system of coupled mass transport and chemical interactions as the source term, which makes the problem non-linear. It is demonstrated that a negligible effect on radiolysis species concentrations change due to the consideration of source term. The crud thickness and porosity effect on the concentration distributions are notably observed. In general, higher concentration starts from the intersection of the heating surface with the chimney wall from the beginning and it reaches the equilibrium state within tens of seconds. The concentration profiles of the radiolysis species H{sub 2}, O{sub 2}, and H{sub 2}O{sub 2} can be directly related to corrosion issues. The direct application of this study to nuclear engineering research is to aid in the design of reactors with higher performance without experiencing an Axial Offset Anomaly (AOA), an unexpected measured shift in axial power distribution from predicted values.

  11. Heat transfer characteristics of current primary packaging systems for pharmaceutical freeze-drying.

    Science.gov (United States)

    Hibler, Susanne; Gieseler, Henning

    2012-11-01

    In the field of freeze-drying, the primary packaging material plays an essential role. Here, the packaging system not only contains and protects the drug product during storage and shipping, but it is also directly involved in the freeze-drying process itself. The heat transfer characteristics of the actual container system influence product temperature and therefore product homogeneity and quality as well as process performance. Consequently, knowledge of the container heat transfer characteristics is of vital importance for process optimization. It is the objective of this review article to provide a summary of research focused on heat transfer characteristics of different container systems for pharmaceutical freeze-drying. Besides the common tubing and molded glass vials and metal trays, more recent packaging solutions like polymer vials, LYOGUARD® trays, syringes, and blister packs are discussed. Recent developments in vial manufacturing are also taken into account.

  12. Not lost in translation: generalization of the primary systems hypothesis to Japanese-specific language processes.

    Science.gov (United States)

    Ueno, Taiji; Saito, Satoru; Saito, Akie; Tanida, Yuki; Patterson, Karalyn; Lambon Ralph, Matthew A

    2014-02-01

    The emergentist-connectionist approach assumes that language processing reflects interaction between primary neural systems (Primary Systems Hypothesis). This idea offers an overarching framework that generalizes to various kinds of (English) language and nonverbal cognitive activities. The current study advances this approach with respect to language in two new and important ways. The first is the provision of a neuroanatomically constrained implementation of the theory. The second is a test of its ability to generalize to a language other than English (in this case Japanese) and, in particular, to a feature of that language (pitch accent) for which there is no English equivalent. A corpus analysis revealed the presence and distribution of typical and atypical accent forms in Japanese vocabulary, forming a quasiregular domain. Consequently, according to the Primary Systems Hypothesis, there should be a greater semantic impact on the processing of words with an atypical pitch accent. In turn, when word meaning is intrinsically less rich (e.g., abstract words), speakers should be prone to regularization errors of pitch accent. We explored these semantic-phonological interactions, first, in a neuroanatomically constrained, parallel-distributed processing model of spoken language processing. This model captured the accent typicality effect observed in nonword repetition in Japanese adults and children and exhibited the predicted semantic impact on repetition of words with atypical accent patterns. Second, also as predicted, in word repetition and immediate serial recall of spoken words, human participants exhibited reduced pitch-accent accuracy and/or slower RT for low imageability words with atypical accent patterns, and they generated accent errors reflecting the more typical accent patterns found in Japanese.

  13. How to achieve optimal organization of primary care service delivery at system level: lessons from Europe.

    Science.gov (United States)

    Pelone, Ferruccio; Kringos, Dionne S; Spreeuwenberg, Peter; De Belvis, Antonio G; Groenewegen, Peter P

    2013-09-01

    To measure the relative efficiency of primary care (PC) in turning their structures into services delivery and turning their services delivery into quality outcomes. Cross-sectional study based on the dataset of the Primary Healthcare Activity Monitor for Europe project. Two Data Envelopment models were run to compare the relative technical efficiency. A sensitivity analysis of the resulting efficiency scores was performed. PC systems in 22 European countries in 2009/2010. Model 1 included data on PC governance, workforce development and economic conditions as inputs and access, coordination, continuity and comprehensiveness of care as outputs. Model 2 included the previous process dimensions as inputs and quality indicators as outputs. There is relatively reasonable efficiency in all countries at delivering as many as possible PC processes at a given level of PC structure. It is particularly important to invest in economic conditions to achieve an efficient structure-process balance. Only five countries have fully efficient PC systems in turning their services delivery into high quality outcomes, using a similar combination of access, continuity and comprehensiveness, although they differ on the adoption of coordination of services. There is a large variation in efficiency levels obtained by countries with inefficient PC in turning their services delivery into quality outcomes. Maximizing the individual functions of PC without taking into account the coherence within the health-care system is not sufficient from a policymaker's point of view when aiming to achieve efficiency.

  14. Work stress of primary care physicians in the US, UK and German health care systems.

    Science.gov (United States)

    Siegrist, Johannes; Shackelton, Rebecca; Link, Carol; Marceau, Lisa; von dem Knesebeck, Olaf; McKinlay, John

    2010-07-01

    Work-related stress among physicians has been an issue of growing concern in recent years. How and why this may vary between different health care systems remains poorly understood. Using an established theoretical model (effort-reward imbalance), this study analyses levels of work stress among primary care physicians (PCPs) in three different health care systems, the United States, the United Kingdom and Germany. Whether professional autonomy and specific features of the work environment are associated with work stress and account for possible country differences are examined. Data are derived from self-administered questionnaires obtained from 640 randomly sampled physicians recruited for an international comparative study of medical decision making conducted from 2005 to 2007. Results demonstrate country-specific differences in work stress with the highest level in Germany, intermediate level in the US and lowest level among UK physicians. A negative correlation between professional autonomy and work stress is observed in all three countries, but neither this association nor features of the work environment account for the observed country differences. Whether there will be adequate numbers of PCPs, or even a field of primary care in the future, is of increasing concern in several countries. To the extent that work-related stress contributes to this, identification of its organizational correlates in different health care systems may offer opportunities for remedial interventions. Copyright 2010 Elsevier Ltd. All rights reserved.

  15. A Case of Primary Central Nervous System Lymphoma Located at Brain Stem in a Child

    Science.gov (United States)

    Kim, Jinho

    2016-01-01

    Primary central nervous system lymphoma (PCNSL) is an extranodal Non-Hodgkin's lymphoma that is confined to the brain, eyes, and/or leptomeninges without evidence of a systemic primary tumor. Although the tumor can affect all age groups, it is rare in childhood; thus, its incidence and prognosis in children have not been well defined and the best treatment strategy remains unclear. A nine-year old presented at our department with complaints of diplopia, dizziness, dysarthria, and right side hemiparesis. Magnetic resonance image suggested a diffuse brain stem glioma with infiltration into the right cerebellar peduncle. The patient was surgically treated by craniotomy and frameless stereotactic-guided biopsy, and unexpectedly, the histopathology of the mass was consistent with diffuse large B cell lymphoma, and immunohistochemical staining revealed positivity for CD20 and CD79a. Accordingly, we performed a staging work-up for systemic lymphoma, but no evidence of lymphoma elsewhere in the body was obtained. In addition, she had a negative serologic finding for human immunodeficient virus, which confirmed the histopathological diagnosis of PCNSL. She was treated by radiosurgery at 12 Gy and subsequent adjuvant combination chemotherapy based on high dose methotrexate. Unfortunately, 10 months after the tissue-based diagnosis, she succumbed due to an acute hydrocephalic crisis. PMID:27867930

  16. Corrosion of carbon steel feeders during dilute chemical decontamination of primary heat transport system of PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Subramanian, H.; Madasamy, P.; Sathyaseelan, V.S.; Krishnamohan, T.V.; Velmurugan, S.; Narasimhan, S.V. [Water and Steam Chemistry Division, BARC Facilities, Kalpakkam, Tamilnadu (India)

    2012-01-15

    Carbon steel feeders in the primary heat transport system of pressurized heavy water reactors (PHWRs) show significant wall thinning due to flow accelerated corrosion (FAC). This is of great concern, as the wear rate in certain locations exceeds the corrosion allowance by design. This necessitates periodic measurement of wall thickness and in some cases even mid course enmasse replacement of feeders. While analyzing the data on wall thicknesses and in arriving at the wall thinning rate during operation of the reactor, sufficient care has to be taken to account for the wall thinning occurring during full system chemical decontamination campaign which is carried out occasionally to reduce dose rates during reactor shut down. Chemical decontamination of primary heat transport system is carried out using a mixture of organic acids at a total concentration of about 0.1 g/L and at 85 C. The results of experiments carried out under simulated conditions for estimating the wall thinning occurring in carbon steel feeder elbow during dilute chemical decontamination are described in this work. The corrosion rates are quantified. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  17. Characteristics of physical activity programs in the Brazilian primary health care system

    Directory of Open Access Journals (Sweden)

    Grace Angélica de Oliveira Gomes

    2014-10-01

    Full Text Available The aim of this study was to describe the characteristics of programs that promote physical activity in the public primary care system by region of Brazil, subject to the presence or absence of multidisciplinary primary care teams (NASF. We conducted a cross sectional and population-based telephone survey of the health unit coordinators from 1,251 health care units. Coordinators were asked about the presence and characteristics of physical activity programs. Four out of ten health units reported having a physical activity intervention program, the most common involving walking groups. Most of the activities were performed in the morning, once or twice a week, and in sessions of 30 minutes or more. Physical education professionals were primarily responsible for directing the activities. Interventions occurred in the health unit itself or in adjacent community spaces. In general, these characteristics were similar between units with or without NASF, but varied substantially across regions. These findings will guide future physical activity policies and programs within primary care in Brazil.

  18. Use of an appendicitis medical information sheet in the pediatric primary care system.

    Science.gov (United States)

    Oyachi, Noboru; Yagasaki, Hideaki; Suzuki, Takeyuki; Higashida, Kosuke; Komai, Takayuki; Hasuda, Norio; Takano, Kunio; Obana, Kazuko

    2016-10-01

    Accurate and prompt diagnosis is required for the primary evaluation of pediatric appendicitis. Among pediatricians and surgeons working in Yamanashi Prefecture, the pediatric appendicitis medical information (PAMI) sheet was edited in April 2011 to reflect the diagnostic results of the pediatric primary and emergency medical service and used as a referral document for surgical consultation to secondary hospitals. The PAMI sheet consisted of sections for history taking, symptoms, physical signs and laboratory findings without a scoring system. For 32 consecutive months starting in April 2011, 59 patients hospitalized for suspected appendicitis were retrospectively reviewed. In particular, a total of 17 referral patients evaluated with the PAMI sheet were assessed in order to evaluate the utility of the form. The pediatric surgeons were able to easily determine patient condition from the PAMI sheet. In total, 13 of 17 patients had appendicitis. According to the physical findings of the 17 studied patients, the judgment of right lower quadrant tenderness (κ = 0.63) and guarding (κ = 1.00) was consistent between the pediatric surgeons and primary attending pediatricians. The PAMI sheet aids in the collection of detailed history and objective data with a high level of accuracy, and provides useful referral diagnostic information to the secondary-level hospitals. © 2016 Japan Pediatric Society.

  19. Implementing a Lean Management System in Primary Care: Facilitators and Barriers From the Front Lines.

    Science.gov (United States)

    Hung, Dorothy; Martinez, Meghan; Yakir, Maayan; Gray, Caroline

    2015-01-01

    Although Lean management techniques are increasingly used in health care to improve quality and reduce costs, lessons about how to successfully implement this approach on the front lines of care delivery are not well documented. In this study, we highlight key facilitators and barriers to implementing Lean among frontline primary care providers. This case study took place at a large, ambulatory care delivery system serving nearly 1 million patients. In-depth interviews were conducted with primary care physicians, staff, and administrators to identify key factors impacting Lean redesigns in primary care. Overall, staff engagement and performance management, sensitivity to the professional values and culture of medicine, and perceived adequacy of organizational resources were critical when introducing Lean changes. Specific drivers of change included empowerment of staff at all levels, visual display of performance metrics, and a culture of innovation and collaboration. Barriers included physician resistance to standardized work, difficulty transferring management responsibilities to non-physician staff, and time and staffing required for participating in improvement efforts. Although Lean offers a new approach to delivering care, the implementation process itself is both complex and crucial to success. Understanding early facilitators and barriers can maximize Lean's, potential to improve health care delivery.

  20. Factors influencing the role of primary care providers as gatekeepers in the Malaysian public healthcare system.

    Science.gov (United States)

    Ang, K T; Ho, B K; Mimi, O; Salmah, N; Salmiah, M S; Noridah, M S

    2014-01-01

    Primary care providers play an important gatekeeping role in ensuring appropriate referrals to secondary care facilities. This cross-sectional study aimed to determine the level, pattern and rate of referrals from health clinics to hospitals in the public sector, and whether the placement of resident family medicine specialist (FMS) had made a significant difference. The study was carried out between March and April in 2012, involving 28 public primary care clinics. It showed that the average referral rate was 1.56% for clinics with resident FMS and 1.94% for those without resident FMS, but it was not statistically significant. Majority of referred cases were considered appropriate (96.1%). Results of the multivariate analysis showed that no prior consultation with senior healthcare provider and illnesses that were not severe and complex were independently associated with inappropriate referrals. Severity, complexity or uncertain diagnosis of patients' illness or injury significantly contributed to unavoidable referrals. Adequate facilities or having more experienced doctors could have avoided 14.5% of the referrals. The low referral rate and very high level of appropriate referrals could indicate that primary care providers in the public sector played an effective role as gatekeepers in the Malaysian public healthcare system.