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Sample records for pwr primary water

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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))

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. 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.

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.).

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  10. 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

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  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. 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)

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  7. 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.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Critical discharge of initially subcooled water through slits. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N; Schrock, V E

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

  11. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  12. 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.

  13. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  14. 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.

  15. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  16. 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)

  17. 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.

  18. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  19. PWR composite materials use. A particular case of safety-related service water pipes

    Energy Technology Data Exchange (ETDEWEB)

    Pays, M.F.; Le Courtois, T

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  20. 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

  1. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  2. 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.

  3. 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.

  4. 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.

  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. 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)

  7. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600

    Science.gov (United States)

    Panter, J.; Viguier, B.; Cloué, J.-M.; Foucault, M.; Combrade, P.; Andrieu, E.

    2006-01-01

    In the present study alloy 600 was tested in simulated pressurised water reactor (PWR) primary water, at 360 °C, under an hydrogen partial pressure of 30 kPa. These testing conditions correspond to the maximum sensitivity of alloy 600 to crack initiation. The resulting oxidised structures (corrosion scale and underlying metal) were characterised. A chromium rich oxide layer was revealed, the underlying metal being chromium depleted. In addition, analysis of the chemical composition of the metal close to the oxide scale had allowed to detect oxygen under the oxide scale and particularly in a triple grain boundary. Implication of such a finding on the crack initiation of alloy 600 is discussed. Significant diminution of the crack initiation time was observed for sample oxidised before stress corrosion tests. In view of these results, a mechanism for stress corrosion crack initiation of alloy 600 in PWR primary water was proposed.

  8. 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.

  9. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. 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...

  11. Once-through steam generator (OTSG) materials and water chemistry. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Pocock, F.J.; Levstek, D.F.

    1974-01-01

    Materials and water chemistry research results associated with the development of the Oconee-1 Reactor steam generator are presented. A summary of water chemistry data acquired during preoperational testing and power operation to date is also included. These data confirm the operational practicality of the nuclear once-through concept using volatile water treatment and high purity condensate demineralized feedwater.

  12. 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)

  13. Effect of water injection on hydrogen generation during severe accident in PWR

    Institute of Scientific and Technical Information of China (English)

    TAO Jun; CAO Xuewu

    2009-01-01

    Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied.The analyses were carried out with different water injection rates at different core damage stages.The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region.Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K,because the core is quenched and reflooded quickly.The water injection at the peak core temperature of 1900 K,the hydrogen generation rate increases at low injection rates of the water,as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate.At peak core temperature of 2100-2300 K,the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core.Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture.Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region.However,hydrogen is generated if water is injected into the molten pool,because steam serves to the crust supporting the molten pool.Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation.Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.

  14. Model of hydrogen-flame interactions with water droplets. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lutz, A.E.

    1982-06-01

    A computer model is developed to study the effects of water droplets on laminar hydrogen deflagrations. The model provides a one-dimensional, transient hydrogen-flame capability using a kinetic chemistry mechanism involving a group of thirteen reactions. Transport equations are solved for mass, thermal energy, and individual species for the gas mixture along with equations for droplet continuity, thermal energy, and size. Calculations show significant cooling of stoichiometric flames for small droplet sizes (20 micron diameters).

  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. Transient analysis of water slug discharge in PWR safety/relief valve piping. [WATSLUG code

    Energy Technology Data Exchange (ETDEWEB)

    Van Duyne, D.A.; Hsieh, J.S.; Shave, D.F.

    1981-01-01

    The sudden discharge of the water loop seal, which is often present upsteam of pressurizer safety and relief valves, creates large momentum and inertia forces on the piping segments downstream of the valve. This paper provides a brief discussion of the commonly available control-volume calculation techniques, a description of the governing equations and a recently developed computer routine (WATSLUG) for their solution, and a review of results calculated using this method for a typical pressurizer safety and relief valve system. 8 refs.

  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. 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

  19. 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.

  20. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  1. 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.

  2. 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.

  3. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in boiling water reactor (BWR) and pressurized water reactor (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, was examined in O/sub 2/ enriched BWR conditions (8 ppm O/sub 2/) and in typical PWR conditions. Cracking susceptibility in BWR conditions is especially sensitive to alpha martensite content and sensitization. Cracking in alpha martensite compounds is intergranular and transgranular and it can not be related to sensitization. Sensitization induces cracking only in creviced conditions (double U-bend specimens) in AISI 304 steels. In creviced conditions OX18H10T steel exhibits cracking in solution annealed, stabilized and sensitized conditions. The sensitized material is most susceptible. Cracking in solution annealed and stabilized OX18H10T steel is intergranular and transgranular. In PWR conditions (O/sub 2/ content 2 ppb) no cracking is observed. (ESA)

  4. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.).

  5. Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H. [Framatome ANP, Inc., Lynchburg, VA (United States); Fyfitch, S. [Framatome ANP, Inc., Lynchburg, VA (United States); Scott, P. [Framatome ANP, SAS, Paris (France); Foucault, M. [Framatome ANP, SAS, Le Creusot (France); Kilian, R. [Framatome ANP, GmbH, Erlangen (Germany); Winters, M. [Framatome ANP, GmbH, Erlangen (Germany)

    2004-03-01

    Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

  6. 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.

  7. 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.

  8. 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.

  9. Influence of localized plasticity on oxidation behaviour of austenitic stainless steels under primary water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cissé, Sarata [CEA Saclay, DEN/DANS/DMN/SEMI, 91191 Gif-sur-Yvette Cedex (France); Laffont, Lydia, E-mail: lydia.laffont@ensiacet.fr [Institut CARNOT, CIRIMAT-ENSIACET, 4 allée Emile Monso, 31030 Toulouse Cedex 4 (France); Lafont, Marie-Christine [Institut CARNOT, CIRIMAT-ENSIACET, 4 allée Emile Monso, 31030 Toulouse Cedex 4 (France); Tanguy, Benoit [CEA Saclay, DEN/DANS/DMN/SEMI, 91191 Gif-sur-Yvette Cedex (France); Andrieu, Eric [Institut CARNOT, CIRIMAT-ENSIACET, 4 allée Emile Monso, 31030 Toulouse Cedex 4 (France)

    2013-02-15

    The sensitivity of precipitation-strengthened A286 austenitic stainless steel to stress corrosion cracking was studied by means of slow-strain-rate tests. First, alloy cold working by low cycle fatigue (LCF) was investigated. Fatigue tests under plastic strain control were performed at different strain levels (Δε{sub p}/2 = 0.2%, 0.5%, 0.8% and 2%) to establish correlations between stress softening and the deformation microstructure resulting from the LCF tests. Deformed microstructures were identified through TEM investigations. The interaction between oxidation and localized deformation bands was also studied and it resulted that localized deformation bands are not preferential oxide growth channels. The pre-cycling of the alloy did not modify its oxidation behaviour. However, intergranular oxidation in the subsurface under the oxide layer formed after exposure to PWR primary water was shown.

  10. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  11. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. 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)

  15. 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

  16. 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.

  17. 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)

  18. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. 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)

  20. 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.

  1. 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)

  2. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  3. 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.

  4. Water Reflooding Effectiveness Assessment for 1 000 MWe PWR under Severe Accident Condition%百万千瓦级压水堆严重事故后再注水的有效性评价

    Institute of Scientific and Technical Information of China (English)

    胡啸; 黄挺; 裴杰; 陈炼

    2015-01-01

    根据现有的设计资料,使用一体化严重事故分析程序 MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08 cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10 kg/s)、中流量(50 kg/s)和大流量(200 kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10 kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200 kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。%The MELCOR1.8.6 code was applied to a severe accident model of a 1 000 MWe PWR which includes primary system,secondary system,passive core cool-ing system and containment system.For the transient case,a small break LOCA with 2 inch (5.08 cm)break at the cold leg concurrent with failure of gravity injection was selected.After the core was damaged due to the failure of gravity inj ection,it was assumed that the coolant was inj ected into the pressure vessel,and then the water reflooding effectiveness was evaluated and analyzed.In this calculation,the coolant injection into reactor core with the small (10 kg/s),medium (50 kg/s)and large (200 kg/s)mass flow rates respectively at 3 different time stages of the severe accident was simulated.The effectiveness of water reflooding was assessed through hydrogen production,radioactive materials released from core,and core temperature.The results show that the mass flow rate above 10 kg/s is believed to be efficient for cooling a 1 000 MWe reactor at the beginning of core damage.However,with the accident devel-oping to core relocation,a large mass flow

  5. 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)

  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. 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.

  8. 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.

  9. Precursor Evolution and Stress Corrosion Cracking Initiation of Cold-Worked Alloy 690 in Simulated Pressurized Water Reactor Primary Water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Toloczko, Mychailo [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Kruska, Karen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.; Bruemmer, Stephen [Pacific Northwest National Laboratory, 622 Horn Rapids Road, P.O. Box 999, Richland, Washington 99352.

    2017-05-22

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 (UNS N06690) 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 either the 21% or 31%CW CLT specimens loaded at their yield stress after ~9,220 hours, 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 10,400 hours of exposure at constant stress intensity, which was 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. 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 discusses their effects on crack initiation in CW alloy 690.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. 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.

  13. Irradiation accelerated corrosion of 316L stainless steel in simulated primary water

    Science.gov (United States)

    Raiman, Stephen S.

    The objective of this work is to understand the effects of irradiation on the corrosion of 316L stainless steel in simulated primary water. 316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to simulated PWR primary water to study the effects of radiation on corrosion. A 3.2 MeV proton beam was transmitted through a 37 microm thick sample that served as a "window" into a corrosion cell containing flowing 320° C water with 3 wppm H2. This design permitted radiolysis and displacement damage to occur on the sample surface in contact with the simulated primary water environment. Samples were irradiated for 4, 12, 24, and 72 hrs at dose rates between 400 and 4000 kGy/s, corresponding to damage rates of 7x10-7 to 7x10-6 dpa/s respectively. The structure and composition of the oxide films were characterized using Raman spectroscopy, STEM, and SEM. Sample areas exposed to direct proton irradiation had inner oxide films that were thinner, more porous, and were deficient in chromium when compared to unirradiated oxides. Outer oxides on irradiated samples exhibited a smaller particle size, and had a significant amount of hematite, which was not found on unirradiated samples. The presence of hematite on irradiated samples indicates an increase in electrochemical potential due to irradiation. Dissolution of chromium-rich spinels due to the elevated potential is identified as a likely mechanism behind the loss of inner oxide chromium. It is suggested that the loss of inner-oxide chromium leads to a less protective inner oxide, and a higher rate of oxide dissolution. Sample areas that were not irradiated, but were exposed to the flow of radiolyzed water, exhibited most of the same phenomena found on irradiated areas including loss of Cr and thinner more porous oxides, indicating that water radiolysis is the primary mechanism. When a sample with a pre-formed oxide was irradiated in the same conditions, the region exposed to radiolyzed

  14. Effect of cold work hardening on stress corrosion cracking of stainless steels in primary water of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Raquet, O.; Herms, E. [CEA/Saclay, DEN/DPC, 91191 Gif sur Yvette Cedex (France); Vaillant, F.; Couvant, T.; Boursier, J.M. [EDF/Les Renardieres, R and D/MMC, 77250 Moret sur Loing (France)

    2004-07-01

    ASSs in primary water of PWR. (authors)

  15. 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)

  16. 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...

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

    Science.gov (United States)

    Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

  5. 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)

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  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. 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

  12. 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

  13. Study of water radiolysis in relation with the primary cooling circuit of pressurized water reactors; Etude sur la radiolyse de l`eau en relation avec le circuit primaire de refroidissement des reacteurs nucleaires a eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Pastina, B

    1997-07-01

    This memorandum shows a fundamental study on the water radiolysis in relation with the cooling primary circuit of PWR type reactors. The water of the primary circuit contains boric acid a soluble neutronic poison and also hydrogen that has for role to inhibit the water decomposition under radiation effect. In the aim to better understand the mechanism of dissolved hydrogen action and to evaluate the impact of several parameters on this mechanism, aqueous solutions with boric acid and hydrogen have been irradiated in a experimental nuclear reactor, at 30, 100 and 200 Celsius degrees. It has been found that, with hydrogen, the water decomposition under irradiation is a threshold phenomenon in function of the ratio between the radiation flux `1` B(n, )`7 Li and the gamma flux. When this ratio become too high, the number of radicals is not sufficient to participate at the chain reaction, and then water is decomposed in O{sub 2} and H{sub 2}O{sub 2} in a irreversible way. The temperature has a beneficial part on this mechanism. The iron ion and the copper ion favour the water decomposition. (N.C.). 83 refs.

  14. 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.

  15. Defect formation in aqueous environment: Theoretical assessment of boron incorporation in nickel ferrite under conditions of an operating pressurized-water nuclear reactor (PWR)

    Science.gov (United States)

    Rák, Zs.; Bucholz, E. W.; Brenner, D. W.

    2015-06-01

    A serious concern in the safety and economy of a pressurized water nuclear reactor is related to the accumulation of boron inside the metal oxide (mostly NiFe2O4 spinel) deposits on the upper regions of the fuel rods. Boron, being a potent neutron absorber, can alter the neutron flux causing anomalous shifts and fluctuations in the power output of the reactor core. This phenomenon reduces the operational flexibility of the plant and may force the down-rating of the reactor. In this work an innovative approach is used to combine first-principles calculations with thermodynamic data to evaluate the possibility of B incorporation into the crystal structure of NiFe2O4 , under conditions typical to operating nuclear pressurized water nuclear reactors. Analyses of temperature and pH dependence of the defect formation energies indicate that B can accumulate in NiFe2O4 as an interstitial impurity and may therefore be a major contributor to the anomalous axial power shift observed in nuclear reactors. This computational approach is quite general and applicable to a large variety of solids in equilibrium with aqueous solutions.

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

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    the reactor core of 1000 MWe PWR performed using MCNP program. The calculation model performed in 9 zones: reactor core, water, baffle, water, barrel, pressure vessel, concrete and the outside air. Determination of the distribution of neutron flux and spectra made to the radial direction to the outside of concrete shield with an accuracy between 10% to 30% in each energy group of 1 and 50 groups. The analysis results of neutron dose rate at the surface of the reactor biological shield of 1000 MWe PWR reactor at full power condition is lower than safety limit value. In terms of neutron radiation exposure, it can be concluded that the two meter thick concrete radiation shielding meets the safety requirements. Key words: PWR NPP, neutron flux, shielding, neutron dose rate, MCNP.

  17. 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.

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. EELS and electron diffraction studies on possible bonaccordite crystals in pressurized water reactor fuel CRUD and in oxide films of alloy 600 material

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jiaxin [Studsvik Nuclear AB, Nykoping (Sweden); Lindberg, Fredrik [Swerea KIMAB AB, Kista (Sweden); Wells, Daniel [Electric Power Research Institute, Charlotte (United States); Bengysson, Bernt [Ringhals AB, Ringhalsverket, Varobacka (Sweden)

    2017-06-15

    Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit) would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS). To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR) cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ⁓2 and similar crystal morphology as the one (bonaccordite) reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of 5 mL H{sub 2}/kg H{sub 2}O, but absent when exposed under 75 mL H{sub 2}/kg H{sub 2}O condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

  6. EELS and electron diffraction studies on possible bonaccordite crystals in pressurized water reactor fuel CRUD and in oxide films of alloy 600 material

    Directory of Open Access Journals (Sweden)

    Jiaxin Chen

    2017-06-01

    Full Text Available Experimental verification of boron species in fuel CRUD (Chalk River Unidentified Deposit would provide essential and important information about the root cause of CRUD-induced power shifts (CIPS. To date, only bonaccordite and elemental boron were reported to exist in fuel CRUD in CIPS-troubled pressurized water reactor (PWR cores and lithium tetraborate to exist in simulated PWR fuel CRUD from some autoclave tests. We have reevaluated previous analysis of similar threadlike crystals along with examining some similar threadlike crystals from CRUD samples collected from a PWR cycle that had no indications of CIPS. These threadlike crystals have a typical [Ni]/[Fe] atomic ratio of ∼2 and similar crystal morphology as the one (bonaccordite reported previously. In addition to electron diffraction study, we have applied electron energy loss spectroscopy to determine boron content in such a crystal and found a good agreement with that of bonaccordite. Surprisingly, such crystals seem to appear also on corroded surfaces of Alloy 600 that was exposed to simulated PWR primary water with a dissolved hydrogen level of 5 mL H2/kg H2O, but absent when exposed under 75 mL H2/kg H2O condition. It remains to be verified as to what extent and in which chemical environment this phase would be formed in PWR primary systems.

  7. 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.

  8. 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.)

  9. 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

  10. 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 ...

  11. Teaching and Learning of 'Water for Agriculture' in Primary Schools ...

    African Journals Online (AJOL)

    Key words: primary school, water for agriculture, rainwater harvesting and ... commercial large-scale farmers. ... The counsel of multiple techniques use is sought from Yin (2008), who ... data was obtained from the curriculum and syllabus documents as well as the .... Enables schools to carry out curriculum based projects,.

  12. UV radiation and primary production in the Antarctic waters

    Digital Repository Service at National Institute of Oceanography (India)

    LokaBharathi, P.A.; Krishnakumari, L.; Bhattathiri, P.M.A.; Chandramohan, D.

    obtained between pp and the above parameters in the Antarctic sub-surface waters determined at discrete depths of 10, 20, 30 and 40 m. However, when the primary productivity values were normalised for PAR, a more negative effect was noticed at the Antarctic...

  13. 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

  14. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers

    NARCIS (Netherlands)

    Gerbens-Leenes, P.W.; Hoekstra, A.Y.; Van der Meer, T.H.

    2007-01-01

    Gerbens-Leenes, P.W., Hoekstra, A.Y., Van der Meer, T.H., 2007. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers. In: proceedings ‘First World Water Sustainability-Renewable Energy Congress and Exhibition’. 25-28 November 2007, Maastricht, the

  15. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers

    NARCIS (Netherlands)

    Gerbens-Leenes, P.W.; Hoekstra, A.Y.; Van der Meer, T.H.

    2007-01-01

    Gerbens-Leenes, P.W., Hoekstra, A.Y., Van der Meer, T.H., 2007. The water footprint of energy consumption: an assessment of water requirements of primary energy carriers. In: proceedings ‘First World Water Sustainability-Renewable Energy Congress and Exhibition’. 25-28 November 2007, Maastricht, the

  16. 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.

  17. 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.

  18. 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.)

  19. 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.).

  20. 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

  1. 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.

  2. 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.

  3. 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.)

  4. 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.

  5. Technical Route and Development of Coolant Circulating Pumps in PWR (Pressurized Water Reactor) Nuclear Power Stations (Ⅱ)%压水堆核电厂冷却剂主循环泵的技术历程和发展(Ⅱ)

    Institute of Scientific and Technical Information of China (English)

    黄经国

    2009-01-01

    本文回顾了压水堆(PWD)核电厂冷却剂主循环泵(简称主泵)从无密封的屏蔽电泵到有轴封泵的发展经历,从核安全要求达成的技术共识,以及世界知名泵厂商在自主化技术背景下各自形成的主泵的技术风格与流派.介绍了主泵技术的改进与创新,以及采用非能动安全系统、优化及简化后的NSSS中.第三代压水堆(PWR)主泵的有关问题.

  6. The problem of optimizing the water chemistry used in the primary coolant circuit of a nuclear power station equipped with VVER reactors under the conditions of longer fuel cycle campaigns and increased capacity of power units

    Science.gov (United States)

    Sharafutdinov, R. B.; Kharitonova, N. L.

    2011-05-01

    It is shown that the optimal water chemistry of the primary coolant circuit must be substantiated while introducing measures aimed at increasing the power output in operating power units and for the project called AES-2006/AES TOI (a typical optimized project of a nuclear power station with enhanced information support). The experience gained from operation of PWR reactors with an elongated fuel cycle at an increased level of power is analyzed. Conditions under which boron compounds are locally concentrated on the fuel rod surfaces (the hideout phenomenon) and axial offset anomaly occurs are enlisted, and the influence of lithium on the hideout in the pores of deposits on the surfaces of fuel assemblies is shown.

  7. 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

  8. EBR-II Primary Tank Wash-Water Alternatives Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Demmer, R. L.; Heintzelman, J. B.; Merservey, R. H.; Squires, L. N.

    2008-05-01

    The EBR-II reactor at Idaho National Laboratory was a liquid sodium metal cooled reactor that operated for 30 years. It was shut down in 1994; the fuel was removed by 1996; and the bulk of sodium metal coolant was removed from the reactor by 2001. Approximately 1100 kg of residual sodium remained in the primary system after draining the bulk sodium. To stabilize the remaining sodium, both the primary and secondary systems were treated with a purge of moist carbon dioxide. Most of the residual sodium reacted with the carbon dioxide and water vapor to form a passivation layer of primarily sodium bicarbonate. The passivation treatment was stopped in 2005 and the primary system is maintained under a blanket of dry carbon dioxide. Approximately 670 kg of sodium metal remains in the primary system in locations that were inaccessible to passivation treatment or in pools of sodium that were too deep for complete penetration of the passivation treatment. The EBR-II reactor was permitted by the Idaho Department of Environmental Quality (DEQ) in 2002 under a RCRA permit that requires removal of all remaining sodium in the primary and secondary systems by 2022. The proposed baseline closure method would remove the large components from the primary tank, fill the primary system with water, react the remaining sodium with the water and dissolve the reaction products in the wash water. This method would generate a minimum of 100,000 gallons of caustic, liquid, low level radioactive, hazardous waste water that must be disposed of in a permitted facility. On February 19-20, 2008, a workshop was held in Idaho Falls, Idaho, to look at alternatives that could meet the RCRA permit clean closure requirements and minimize the quantity of hazardous waste generated by the cleanup process. The workshop convened a panel of national and international sodium cleanup specialists, subject matter experts from the INL, and the EBR-II Wash Water Project team that organized the workshop. The

  9. 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.

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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.)

  19. 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)

  20. 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.).

  1. 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)

  2. 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。

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 75 FR 53267 - National Primary Drinking Water Regulations: Revisions to the Total Coliform Rule; Extension of...

    Science.gov (United States)

    2010-08-31

    ... AGENCY 40 CFR Parts 141 and 142 RIN 2040-AD94 National Primary Drinking Water Regulations: Revisions to... 30 days the public comment period for a proposed National Primary Drinking Water Regulation, the... inquiries, contact Sean Conley, Office of Ground Water and Drinking Water (MC 4607M), U.S....

  8. 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.

  9. 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.

  10. 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....

  11. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  12. 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.

  13. Decreased growth-induced water potential: A primary cause of growth inhibition at low water potentials

    Energy Technology Data Exchange (ETDEWEB)

    Nonami, Hiroshi [Ehime Univ., Matsuyama (Japan); Wu, Yajun; Boyer, J.S. [Univ. of Delaware, Lewes, DE (United States)

    1997-06-01

    Cell enlargement depends on a growth-induced difference in water potential to move water into the cells. Water deficits decrease this potential difference and inhibit growth. To investigate whether the decrease causes the growth inhibition, pressure was applied to the roots of soybean seedlings and the growth and potential difference were monitored in the stems. In water-limited plants, the inhibited stem growth increased when the roots were pressurized and it reverted to the previous rate when the pressure was released. The pressure around the roots was perceived as an increased turgor in the stem in small cells next to the xylem, but not in outlying cortical cells. This local effect implied that water transport was impeded by the small cells. The diffusivity for water was much less in the small cells than in the outlying cells. The small cells thus were a barrier that caused the growth-induced potential difference to be large during rapid growth, but to reverse locally during the early part of a water deficit. Such a barrier may be a frequent property of meristems. Because stem growth responded to the pressure-induced recovery of the potential difference across this barrier, we conclude that a decrease in the growth-induced potential difference was a primary cause of the inhibition.

  14. 75 FR 30401 - National Primary Drinking Water Regulations; Announcement of the Results of EPA's Review of...

    Science.gov (United States)

    2010-06-01

    ... AGENCY RIN 2040-AE90 National Primary Drinking Water Regulations; Announcement of the Results of EPA's Review of Existing Drinking Water Standards and Request for Public Comment and/or Information on Related... the public comment period for the National Primary Drinking Water Regulations; Announcement of...

  15. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  16. 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.)

  17. 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

  18. 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).

  19. 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.

  20. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. 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.

  2. 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.

  3. 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.

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

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Full Text Available AP1000 adalah reaktor daya PWR maju dengan daya listrik 1154 MW yang didesain berdasarkan kinerja teruji dari desain PWR lain oleh Westinghouse. Untuk mempersiapkan peran Pusat Teknologi Reaktor dan Keselamatan Nuklir sebagai suatu Technical Support Organization (TSO dalam hal verifikasi keselamatan, telah dilakukan kegiatan verifikasi keselamatan untuk AP1000 yang dimulai dengan verifikasi kecelakaan kegagalan pendingin sekunder. Kegiatan dimulai dengan pemodelan fitur keselamatan teknis yaitu sistem pendinginan teras pasif yang terdiri dari sistem Passive Residual Heat Removal (PRHR, tangki core makeup tank (CMT, dan tangki In-containment Refueling Water Storage Tank (IRWST. Kecelakaan kegagalan pendingin sekunder yang dipilih adalah hilangnya aliran air umpan ke salah satu pembangkit uap yang disimulasikan menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Tujuan analisis adalah untuk memperoleh sekuensi perubahan parameter termohidraulika reaktor akibat kecelakaan dimana hasil analisis yang diperoleh divalidasi dan dibandingkan dengan hasil analisis menggunakan program perhitungan LOFTRAN di dalam dokumen desain keselamatan AP1000. Hasil verifikasi menunjukkan bahwa kejadian hilangnya suplai air umpan tidak berdampak pada kerusakan teras, sistem pendingin reaktor, maupun sistem sekunder. Penukar kalor PRHR telah terverifikasi kemampuannya dalam membuang kalor peluruhan teras setelah trip reaktor. Hasil validasi dengan dokumen pembanding menunjukkan kesesuaian pada sebagian besar parameter termohidraulika. Secara umum, model PWR maju yang dilengkapi dengan sistem pendinginan teras ciri pasif yang telah dikembangkan tetap selamat ketika terjadi kecelakaan kehilangan aliran pendingin sekunder. Kata kunci: Verifikasi, hilangnya aliran air umpan, AP1000   AP1000 is a PWR power reactor with 1154 MW of electrical power that is designed based on the proven performance of the other Westinghouse PWR designs. To prepare the role of Center for

  5. Water-soluble primary amine compounds in rural continental precipitation

    Science.gov (United States)

    Gorzelska, Krystyna; Galloway, James N.; Watterson, Karen; Keene, William C.

    Procedures for collecting, storing and analysing precipitation samples for organic nitrogen studies were developed. These procedures preserve chemical integrities of the species of interest, allow for up to 3 months storage and quantitative determination of water-soluble primary amine compounds, with the overall error at the 2 nM detection limit of less than 30%. This methodology was applied to study amino compounds in precipitation samples collected over a period of one year in central Virginia. Nitrogen concentrations of 13 amino acids and 3 aliphatic amines were summed to calculate the total amine nitrogen (TAN). The concentration of TAN ranged from below our detection level to 6658 nM, and possibly reflected a seasonal variation in the source strength of the atmospheric amines. Overall, the most commonly occurring amino compounds were methyl amine, ethyl amine, glutamic acid, glycine and serine. On average, the highest overall contribution to the TAN came from arginine, asparagine, glutamine, methyl amine, serine and alanine. However, large qualitative and quantitative variations observed among samples warrant caution in interpretation and application of the averaged values. TAN in Charlottesville precipitation contributed from less than 1 to ca 10% of the ammonium nitrogen level. However, our estimates show that amino compounds may contribute significantly to reduced nitrogen budget in precipitation in remote regions.

  6. 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.

  7. 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.

  8. The effect of water resistance in the primary coating by silane coupling agents and amine synergists

    Science.gov (United States)

    Kim, Min-Jeong; Min, Kyoungbeom; Kim, Sanghwan; Yoon, Jung-Woo; Kwak, Seungjo; Lee, Ji-hye; Oh, Sungkoog

    2008-11-01

    In order to clarify the relationship between the water resistance and compositional additives in the primary coating for optical fibers, we prepared various primary coatings that contained silane coupling agents in combination with amine synergists. We observed the appearance of the interface between glass and primary coating after soaking in water at 65° for 30 days. Water resistance was found to be heavily influenced by the content and type of silane coupling agents and amine synergists and their trends are reported.

  9. 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.

  10. 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.

  11. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  12. 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.

  13. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  14. Dynamic Simulations of Primary Frequency Regulation for Pressurized Water Reactor Nuclear Power Generation Units%压水堆核电机组一次调频动态仿真

    Institute of Scientific and Technical Information of China (English)

    唐贞鹏; 陈世和; 伍宇忠; 王鹏飞; 方华伟; 赵福宇

    2013-01-01

    Along with rapid development of nuclear power in China, installed capacity of nuclear power units is increased unceasingly. Due to various reasons at present the nuclear power units in China are operated in basic-load running mode and do not participate in power grid frequency regulation, and it greatly affects the control of grid frequency. In allusion to this phenomenon, taking the Daya Bay nuclear power plant as research object, the feasibility of nuclear power units participating primary frequency regulation (PFR) of power grid is researched. A nonlinear time-varying dynamic model of overall nuclear power plant composed of pressurized water reactor (PWR) nuclear power units is established, and using FORTRAN language a dynamic calculation software is programmed. Then the dynamic calculation program is compiled as dynamic link library (DLL) files and embedded into Matlab/Simulink simulation platform in the form of S-function, and then in Matlab/Simulink environment a PFR simulation platform of PWR nuclear power units is built, and dynamic simulation of PFR is performed. Simulation results show that under current design of PWR nuclear power units it is feasible for PWR nuclear power units to participate PFR in the viewpoint of economy and security.%随着我国核电的快速发展,核电装机不断增加。但是目前我国核电机组由于各种原因,始终是以基本负荷模式运行,不参与电网调频,这对电网频率的控制带来了巨大冲击。针对此问题,以大亚湾核电站为研究对象,对核电机组参与电网一次调频的可行性进行了研究。建立了压水堆核电机组全电厂的非线性时变动态模型,用 FORTRAN 语言编写了动态计算程序。然后把动态计算程序编译为动态链接库文件,并通过以S函数的方式接入Matlab/Simulink仿真平台,在 Matlab/Simulink 中进行了核电机组一次调频仿真平台搭建和一次调频动态仿真。仿真结果表明,在

  15. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  16. 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)

  17. 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.

  18. 40 CFR Appendix III to Part 265 - EPA Interim Primary Drinking Water Standards

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 25 2010-07-01 2010-07-01 false EPA Interim Primary Drinking Water..., STORAGE, AND DISPOSAL FACILITIES Pt. 265, App. III Appendix III to Part 265—EPA Interim Primary Drinking Water Standards Parameter Maximum level (mg/l) Arsenic 0.05 Barium 1.0 Cadmium 0.01 Chromium...

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. 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)

  7. 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)

  8. Primary Datasets for Case Studies of River-Water Quality

    Science.gov (United States)

    Goulder, Raymond

    2008-01-01

    Level 6 (final-year BSc) students undertook case studies on between-site and temporal variation in river-water quality. They used professionally-collected datasets supplied by the Environment Agency. The exercise gave students the experience of working with large, real-world datasets and led to their understanding how the quality of river water is…

  9. Primary Datasets for Case Studies of River-Water Quality

    Science.gov (United States)

    Goulder, Raymond

    2008-01-01

    Level 6 (final-year BSc) students undertook case studies on between-site and temporal variation in river-water quality. They used professionally-collected datasets supplied by the Environment Agency. The exercise gave students the experience of working with large, real-world datasets and led to their understanding how the quality of river water is…

  10. The water footprint of bioenergy and other primary energy carriers

    NARCIS (Netherlands)

    Gerbens-Leenes, Winnie; Hoekstra, Arjen Ysbert; van der Meer, Theodorus H.

    2008-01-01

    Freshwater is essential for life on earth, not only for basic human needs such as food, fibre and drinking water, but also for a healthy environment. In the near future, important challenges are to meet basic needs and to ensure that the extraction of water does not affect freshwater ecosystems. At

  11. Water safety: one of the primary objectives of our time

    Directory of Open Access Journals (Sweden)

    Carlo Collivignarelli

    2017-01-01

    Full Text Available This article discusses the benefits of an innovative approach to the problem of water security introduced by WHO in 2004, through the establishment of the Water Safety Plan (WSP. It was recently included in Commission Directive (EU 2015/1787 - October 6, 2015 - the implementation of which is expected in the EU countries by 27 October 2017. The WSP is the most effective means of consistently ensuring the safety of a drinking water supply. The method is based on the use of a comprehensive risk assessment and risk management approach that involves all steps in water supply from catchment to consumer. The knowledge acquired by some experiences of WSP application, both in Italy and in countries with limited resources, is proving the effectiveness of the model as the best way to manage drinking water systems and protect public health.

  12. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  13. 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)

  14. 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.

  15. 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.

  16. 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)

  17. Primary productivity in nearshore waters of Thal, Maharashtra coast

    Digital Repository Service at National Institute of Oceanography (India)

    Varshney, P.K.; Nair, V.R.; Abidi, S.A.H.

    Primary productivity off Thal, Maharashtra, India was evaluated at 3 stations during Feb. 1980 to Jan. 1981. The area was quite turbid and the euphotic zone never exceeded 2.5 m. Column production ranged from 0.69 to 605.21 mg C.m/2.d/2 (av. 78.2 mg...

  18. Computational analysis for prediction of pressure of PWR presurizer undertransient conditions

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A computer model has been developed for prediction of the pressure in thepressurizer undertransient conditions. In the model three separate thermodynamic regions which arenot required to be inthermal equilibrium have been considered. The mathematical model derived from the general conservation equations includesall of theimportant thermal-hydraulics phenomena occurring in the pressurizer,i.e., stratificationof the hot water andincoming cold water, bulk flashing and condensation, wall condensation, andinterfacial heat and masstransfer, etc. The bubble rising and rain-out models are developed to describe bulkflashing andcondensation, respectively. To obtain the wall condensation rate, a one-dimensionalheat conductionequation is solved by the pivoting method. The presented model will predict thepressure-time behaviorof a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreementwithavailable data on the CHASHMA nuclear power plant's pressurizer performance.

  19. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    Institute of Scientific and Technical Information of China (English)

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  20. Elementary analysis of the corrosion products in the nuclear power plants` primary water; Ydinvoimalaitosveden korroosiotuotteiden alkuaineanalytiikka

    Energy Technology Data Exchange (ETDEWEB)

    Rosenberg, R.

    1995-02-01

    The study, based on a literature survey, concentrates on the sampling and analysis methods of the dissolved or colloidal transition metals: iron, manganese, nickel and cobalt in the primary coolant waters of the nuclear power plants. (73 refs.).

  1. Synthesis of highly functionalized pyrroles from primary amines and activated acetylenes in water

    Institute of Scientific and Technical Information of China (English)

    S.Zahra Sayyed-Alangi; Zinatossadat Hossaini; Faramarz Rostami-Charati

    2012-01-01

    Stable derivatives of pyrroles were prepared using multicomponent reactions of dialkyl acetylenedicarboxylate,primary amine and propiolate in the presence of N-methylimidazole in water at room temperature in good yields.

  2. 压水堆核电厂冷却剂主循环泵的技术历程和发展(Ⅰ)%Technical Route and Development of Coolant Circulating Pumps in PWR(Pressurized Water Reactor) Nuclear Power Stations(Ⅰ)

    Institute of Scientific and Technical Information of China (English)

    黄经国

    2009-01-01

    本文同顾了压水堆(PWR)核电厂冷却剂主循环泵(简称主泵)从无密封的屏蔽电泵到有轴封泵的发展经历,从核安全要求达成的技术共识,以及世界知名泵厂商在自主化技术背景下各自形成的主泵的技术风格与流派.介绍了主泵技术的改进与创新,以及采用非能动安全系统、优化及简化后的NSSS中,第三代压水堆(PWR)主泵的有关问题.

  3. The aquaporin-2 water channel in autosomal dominant primary nocturnal enuresis.

    NARCIS (Netherlands)

    Deen, P.M.T.; Dahl, N.; Caplan, M.J.

    2002-01-01

    PURPOSE: Nocturnal enuresis is one of the most common diagnoses in a pediatric clinic. Recently, linkage analysis revealed a 2-point lod score of 4.2 in 6 families with dominant primary nocturnal enuresis around the aquaporin-2 (AQP2) water channel locus. Since primary nocturnal enuresis is ameliora

  4. 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.)

  5. 40 CFR 142.19 - EPA review of State implementation of national primary drinking water regulations for lead and...

    Science.gov (United States)

    2010-07-01

    ... national primary drinking water regulations for lead and copper. 142.19 Section 142.19 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) WATER PROGRAMS (CONTINUED) NATIONAL PRIMARY DRINKING... implementation of national primary drinking water regulations for lead and copper. (a) Pursuant to the...

  6. 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.

  7. 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).

  8. 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)

  9. 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.

  10. 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.

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. 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.

  14. 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)

  15. 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.

  16. 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.

  17. 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)

  18. 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.

  19. Primary acoustic signal structure during free falling drop collision with a water surface

    Energy Technology Data Exchange (ETDEWEB)

    Chashechkin, Yu. D., E-mail: chakin@ipmnet.ru; Prokhorov, V. E., E-mail: prohorov@ipmnet.ru [Russian Academy of Sciences, Ishlinskii Institute for Problems in Mechanics (Russian Federation)

    2016-04-15

    Consistent optical and acoustic techniques have been used to study the structure of hydrodynamic disturbances and acoustic signals generated as a free falling drop penetrates water. The relationship between the structures of hydrodynamic and acoustic perturbations arising as a result of a falling drop contacting with the water surface and subsequent immersion into water is traced. The primary acoustic signal is characterized, in addition to stably reproduced features (steep leading edge followed by long decay with local pressure maxima), by irregular high-frequency packets, which are studied for the first time. Reproducible experimental data are used to recognize constant and variable components of the primary acoustic signal.

  20. Simplified, rapid, and inexpensive estimation of water primary productivity based on chlorophyll fluorescence parameter Fo.

    Science.gov (United States)

    Chen, Hui; Zhou, Wei; Chen, Weixian; Xie, Wei; Jiang, Liping; Liang, Qinlang; Huang, Mingjun; Wu, Zongwen; Wang, Qiang

    2017-04-01

    Primary productivity in water environment relies on the photosynthetic production of microalgae. Chlorophyll fluorescence is widely used to detect the growth status and photosynthetic efficiency of microalgae. In this study, a method was established to determine the Chl a content, cell density of microalgae, and water primary productivity by measuring chlorophyll fluorescence parameter Fo. A significant linear relationship between chlorophyll fluorescence parameter Fo and Chl a content of microalgae, as well as between Fo and cell density, was observed under pure-culture conditions. Furthermore, water samples collected from natural aquaculture ponds were used to validate the correlation between Fo and water primary productivity, which is closely related to Chl a content in water. Thus, for a given pure culture of microalgae or phytoplankton (mainly microalgae) in aquaculture ponds or other natural ponds for which the relationship between the Fo value and Chl a content or cell density could be established, Chl a content or cell density could be determined by measuring the Fo value, thereby making it possible to calculate the water primary productivity. It is believed that this method can provide a convenient way of efficiently estimating the primary productivity in natural aquaculture ponds and bringing economic value in limnetic ecology assessment, as well as in algal bloom monitoring.

  1. 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

  2. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  3. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  4. 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.

  5. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  6. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR 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)

    2010-10-15

    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  7. Intraocular pressure fluctuation after water drinking test in primary angle-closure glaucoma and primary open-angle glaucoma

    Directory of Open Access Journals (Sweden)

    Yi-Chieh Poon

    2016-01-01

    Full Text Available Context: Only a few studies have assessed intraocular pressure (IOP changes during the water drinking test (WDT in patients with primary angle-closure glaucoma (PACG. Aims: The aim of this study is to investigate IOP changes during WDT in patients with PACG versus primary open-angle glaucoma (POAG. Settings and Design: This was a prospective and single tertiary center study. Materials and Methods: PACG and POAG patients (n = 15 each without prior glaucoma surgery were enrolled and subjected to WDT, wherein they consumed an amount of water proportional to their body weight within 10 min. IOP was measured at baseline and every 15 min for 1 h after water intake. Statistical Analysis Used: Intergroup comparisons were performed using Mann–Whitney U-test for continuous variables and Chi-square test for categorical variables. Wilcoxon signed-ranks test was used for comparisons of IOP before and after water intake in the two groups. Regression analysis was used to determine factors associated with IOP fluctuations during WDT. Results: IOP changes over 1 h after water intake showed no significant differences between groups. The mean maximum fluctuation from baseline was 3.61 ± 2.49 and 3.79 ± 1.91 mmHg, respectively, in the PACG and POAG groups. The mean peak IOP was 19.17 ± 4.32 and 19.87 ± 3.44 mmHg in the PACG and PAOG groups, respectively. The axial length and anterior chamber depth showed no correlations with IOP fluctuations. Conclusions: We found similar IOP fluctuation curves and peak IOP values in both PACG and POAG patients subjected to WDT. These findings suggest that WDT is a useful test to induce IOP peaks in both POAG and PACG patients.

  8. Primary Approach to Water (Natural Monopoly) Industry Remodeling

    Institute of Scientific and Technical Information of China (English)

    XiaChengxiang

    2005-01-01

    Natural monopoly, because of its spontaneous or natural characteristics, most have some mysterious causes of reasonableness. Thus, the regulations to its efficiency loss would have a different way compared to other monopolies. That the characteristics of natural monopoly, in the case of water industry, are that the infrastructure investments are very large, most of which are used to build the transportation system? The webs for transporting their products to their customers, and the products are identity goods or services. By examining the characteristics of natural monopoly, this paper proposed a way to break up and remodel the industry of """"""""natural monopoly'. The main clue of remodeling is that the govemments, who represent the public and who ha ve the power to control over public resources, should build and maintain a public web platform for the goods' transportation uses, and break up the barrier of the entry so as to produce a market-oriented competitive structure, The running model and the condition of remodeling are put forward and the cost-revenue analysis of the operation is briefly under consideration.

  9. Potable water quality monitoring of primary schools in Magura district, Bangladesh: children's health risk assessment.

    Science.gov (United States)

    Rahman, Aminur; Hashem, Abul; Nur-A-Tomal, Shahruk

    2016-12-01

    Safe potable water is essential for good health. Worldwide, school-aged children especially in the developing countries are suffering from various water-borne diseases. In the study, drinking water supplies for primary school children were monitored at Magura district, Bangladesh, to ensure safe potable water. APHA standard analytical methods were applied for determining the physicochemical parameters of the water samples. For determination of the essential physicochemical parameters, the samples were collected from 20 randomly selected tube wells of primary schools at Magura. The metal contents, especially arsenic (As), iron (Fe), and manganese (Mn), in the water samples were analyzed by atomic absorption spectroscopy. The range of physicochemical parameters found in water samples were as follows: pH 7.05-9.03, electrical conductivity 400-2340 μS/cm, chloride 10-640 mg/L, hardness 200-535 mg/L as CaCO3, and total dissolved solids 208-1216 mg/L. The level of metals in the tube well water samples were as follows: As 1 to 55 μg/L, Fe 40 to 9890 μg/L, and Mn 10 to 370 μg/L. Drinking water parameters of Magura district did not meet the requirement of the World Health Organization drinking water quality guideline, or the Drinking Water Quality Standards of Bangladesh.

  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. Simulation of the occupational radiation dose caused by contamination of primary circuit media in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

    The occupational radiation exposure of workers in NPPs during overall maintenance and refueling inspections and decommissioning is determined by numerous parameters. Radiation exposure caused by contamination of components may be minimised by the chemical operation mode and by applying systematic decontamination techniques. Data on occupational exposure in German NPPs as well as information about the radionuclide concentration in the coolant are available. The generic 3D model of the primary circuit presented is based on the analysis of technical documentation of German PWRs. Tasks are modeled as a combination of retention times at related local positions in the surroundings of work areas. The generic model allows the calculation of the resulting occupational doses generated by definable jobs and tasks. The KWU/Siemens- PWR generations are characterised by nuclide vectors, the thickness of shielding, and the material composition of components. It was possible to show that for a pre-Konvoi plant, the calculated occupational dose caused by a specific working task is close to measurements.

  12. 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.

  13. Advanced primary treatment of waste water using a bio-flocculation-adsorption sedimentation process

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, W.; Ting, Y.P.; Chen, J.P.; Xing, C.H. [National Univ., Singapore (Singapore). Dept. of Chemical and Environmental Engineering; Shi, S.Q. [Tsinghua Univ., Beijing (China). Dept. of Environmental Science and Engineering

    2000-07-01

    An advanced primary treatment process for a municipal waste water was systematically studied, using a bio-flocculation-adsorption, sedimentation and stabilization process (BSS). It was shown that the organic removal efficiency was higher than that of the traditional primary treatment processes but lower than that of the traditional secondary treatment processes. Both adsorption and bio-flocculation played an important role in the removal of pollutants. The activated sludge within the bio-flocculation-adsorption tank could be considered a bio-flocculent which improved the quality of the effluent from the primary treatment process. As the effluent of the BSS process did not meet the requirements for a typical secondary effluent, the process may be regarded as an advanced (or enhanced) primary treatment process, suitable for waste water containing a high concentration of suspended solids and colloidal particles. (orig.)

  14. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  15. 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.

  16. 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)

  17. 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

  18. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

    2017-03-15

    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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.

  4. Primary collector wall local temperature fluctuations in the area of water-steam phase boundary

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O.; Klinga, J.; Simo, T. [Energovyzkum Ltd., Brno (Switzerland)

    1995-12-31

    A limited number of temperature sensors could be installed at the primary collector surface in the area of water - steam phase boundary. The surface temperatures as well WWER 440 steam generator process data were measured and stored for a long time and off-line evaluated. Selected results are presented in the paper. (orig.). 2 refs.

  5. Northeast Atlantic Late Quaternary planktic Foraminifera as primary productivity and water mass indicators

    NARCIS (Netherlands)

    Kreveld, van S.A.

    1996-01-01

    Primary productivity and water mass reconstructions based on planktic Foraminifera reveal distinct interglacial/glacial variations for the past 208 ka in a mid-latitude Northeast Atlantic piston core. Average total planktic foraminiferal absolute frequencies and accumulation rates, which are

  6. 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)

  7. 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)

  8. 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)

  9. Conventional Treatment of Surface Water Using Moringa Oleifera Seeds Extract as a Primary Coagulant

    Directory of Open Access Journals (Sweden)

    Suleyman A. Muyibi, Ahmed Hissein M Birima, Thamer A. Mohammed

    2012-10-01

    Full Text Available The present study involved the use of a model pilot scale water treatment plant to treat turbid surface water from a stream using processed Moringa oleifera seed with 25 % w/w oil extracted as primary coagulant. The water treatment plant was made up of four unit operations: coagulation, flocculation, sedimentation, and filtration (rapid sand filter. Test runs were carried out for three hours per run over a three-month period with turbidities ranging from 18 to 261 NTU. The turbidity, pH, and alkalinity as well as the filter head loss were measured every 30 minutes during the experimental runs. Average turbidity removal of up to 96 % at an effective doses of 20 and 30 mg/l of oil extracted M. oleifera for low (< 50 NTU and moderate turbidity (< 100 NTU water respectively was observed doses 50 – 80 mg/l for high turbidity (> 100 NTU water. M. oleifera seed extract was found to have no significant effect on pH or alkalinity of the water. The residual turbidities measured during most of the test runs satisfied the Malaysian Guideline for Drinking Water Supplies. Key Words: Moringa oleifera, primary coagulant, coagulation, pilot plant, filtration.

  10. 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.

  11. Is copper an inhibiting factor for primary production in the upwelling waters of Cabo Frio?

    Directory of Open Access Journals (Sweden)

    Diniz Antônia G.

    2003-01-01

    Full Text Available The goal of the present work was to test the hypothesis that inorganic species of copper may inhibit primary production in upwelling waters from the Cabo Frio region. Water samples were collected from four different depths up to 50 m at three stations of Cabo Frio and nutrient concentration, pH, dissolved oxygen, alkalinity, seston, chlorophyll a, primary production, NPP, copper speciation, Cu(II and Cu(I and complexing capacity were determined. The results indicate that the rate of photosynthesis in nutrient rich waters with higher copper content and lower complexing capacity is significantly reduced. Cu(I made up 2-15% of the total copper, however, there was no evidence to suggest that the photochemical production of Cu(I affects NPP.

  12. 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.)

  13. 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)

  14. Water consumption in artificial desert oasis based on net primary productivity

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    Analysis of the water consumption is the basis for water allocation in oasis. However, the method of estimating oasis water consumption remains a great challenge. Based on net primary productivity (NPP) and the transpiration coefficient, a vegetation water consumption model was developed to estimate the water consumption in desert oasis in ERDAS environment. Our results demonstrated that the ecosystem in the middle reaches of the Heihe oasis consumed water of 18.41×108-21.9×108 m3 for irrigation. Without taking precipitation into account, the water consumption in farmland accounted for 77.1%-77.8% (or about 13.97×108-16.84×108 m3) of the oasis vegetation water consumption and in the farmland protection system accounting for 22%. The growing period precipitation in desert environments is about 7.02×108 m3, and the total annual precipitation is about 8.29×108 m3. The modeled water consumption of desert vegetation, however, is about 4.57×108 m3, equivalent to only 65% of the growing period precipitation or 55% of the total annual precipitation. The modeled value equals to the cumulative precipitation of greater than 5 mm, which is defined as the effective precipitation in arid desert.

  15. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  16. 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.

  17. Isolation of the etiological agent of primary amoebic meningoencephalitis from artificially heated waters

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, A.R.; Tyndall, R.L.; Coutant, C.C.; Willaert, E.

    1977-12-01

    To determine whether artificial heating of water by power plant discharges facilitates proliferation of the pathogenic free-living amoebae that cause primary amoebic meningoencephalitis, water samples (250 ml) were taken from discharges within 3,000 feet (ca. 914.4 m) of power plants and were processed for amoeba culture. Pathogenic Naegleria fowleri grew out of water samples from two of five lakes and rivers in Florida and from one of eight man-made lakes in Texas. Pathogenic N. fowleri did not grow from water samples taken from cooling towers and control lakes, the latter of which had no associated power plants. The identification of N. fowleri was confirmed by pathogenicity in mice and by indirect immunofluorescence analyses, by using a specific antiserum.

  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. The Life-Cycle Costs of School Water, Sanitation and Hygiene Access in Kenyan Primary Schools

    OpenAIRE

    Kelly T. Alexander; Alex Mwaki; Dorothy Adhiambo; Malaika Cheney-Coker; Richard Muga; Freeman, Matthew C

    2016-01-01

    Water, Sanitation and Hygiene (WASH) programs in schools can increase the health, dignity and comfort of students and teachers. Understanding the costs of WASH facilities and services in schools is one essential piece for policy makers to utilize when budgeting for schools and helping to make WASH programs more sustainable. In this study we collected data from NGO and government offices, local hardware shops and 89 rural primary schools across three Kenyan counties. Current expenditures on WA...

  20. 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.

  1. Solution of a benchmark set problems for BWR and PWR reactors with UO{sub 2} and MOX fuels using CASMO-4; Solucion de un Conjunto de Problemas Benchmark para Reactores BWR y PWR con Combustible UO{sub 2} y MOX Usando CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G. [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: mike_ipn_esfm@hotmail. com

    2007-07-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO{sub 2}) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  2. 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.

  3. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Directory of Open Access Journals (Sweden)

    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  4. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Science.gov (United States)

    2013-09-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews...

  5. Wintertime Arctic Ocean sea water properties and primary marine aerosol concentrations

    Directory of Open Access Journals (Sweden)

    J. Zábori

    2012-11-01

    Full Text Available Sea spray aerosols are an important part of the climate system through their direct and indirect effects. Due to the diminishing sea ice, the Arctic Ocean is one of the most rapidly changing sea spray aerosol source areas. However, the influence of these changes on primary particle production is not known.

    In laboratory experiments we examined the influence of Arctic Ocean water temperature, salinity, and oxygen saturation on primary particle concentration characteristics. Sea water temperature was identified as the most important of these parameters. A strong decrease in sea spray aerosol production with increasing water temperature was observed for water temperatures between −1°C and 9°C. Aerosol number concentrations decreased from at least 1400 cm−3 to 350 cm−3. In general, the aerosol number size distribution exhibited a robust shape with one mode close to dry diameter Dp 0.2 μm with approximately 45% of particles at smaller sizes. Changes in sea water temperature did not result in pronounced change of the shape of the aerosol size distribution, only in the magnitude of the concentrations. Our experiments indicate that changes in aerosol emissions are most likely linked to changes of the physical properties of sea water at low temperatures. The observed strong dependence of sea spray aerosol concentrations on sea water temperature, with a large fraction of the emitted particles in the typical cloud condensation nuclei size range, provide strong arguments for a more careful consideration of this effect in climate models.

  6. Wintertime Arctic Ocean sea water properties and primary marine aerosol concentrations

    Directory of Open Access Journals (Sweden)

    J. Zábori

    2012-06-01

    Full Text Available Sea spray aerosols are an important part of the climate system through their direct and indirect effects. Due to the diminishing sea ice, the Arctic Ocean is one of the most rapidly changing sea spray aerosol source areas. However, the influence of these changes on primary particle production is not known.

    In laboratory experiments we examined the influence of Arctic Ocean water temperature, salinity and oxygen saturation on primary particle concentration characteristics. Sea water temperature was identified as the most important of these parameters. A strong decrease in sea spray aerosol production with increasing water temperature was observed for water temperatures between −1 °C and 9 °C. Aerosol number concentrations decreased from at least 1400 cm−3 to 350 cm−3. In general, the aerosol number size distribution exhibited a robust shape with one mode close to Dp 0.2 μm with approximately 45% of particles at smaller sizes. Changes in sea water temperature did not result in pronounced change of the shape of the aerosol size distribution, only in the magnitude of the concentrations. Our experiments indicate that changes in aerosol emissions are most likely linked to changes of the physical properties of sea water at low temperatures. The observed strong dependence of sea spray aerosol concentrations on sea water temperature, with a large fraction of the emitted particles in the typical cloud condensation nuclei size range, provide strong arguments for a more careful consideration of this effect in climate models.

  7. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  8. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  9. Primary production of coral ecosystems in the Vietnamese coastal and adjacent marine waters

    Science.gov (United States)

    Tac-An, Nguyen; Minh-Thu, Phan; Cherbadji, I. I.; Propp, M. V.; Odintsov, V. S.; Propp, L. H.

    2013-11-01

    Coral reef ecosystems in coastal waters and islands of Vietnam have high primary production. Average gross primary production (GPP) in coral reef waters was 0.39 g C m-2 day-1. GPP of corals ranged from 3.12 to 4.37 g C m-2 day-1. GPP of benthic microalgae in coral reefs ranged from 2 to 10 g C m-2 day-1. GPP of macro-algae was 2.34 g C m-2 day-1. Therefore, the total of GPP of whole coral reef ecosystems could reach 7.85 to 17.10 g C m-2 day-1. Almost all values of the ratio of photosynthesis to respiration in the water bodies are higher than 1, which means these regions are autotrophic systems. Wire variation of GPP in coral reefs was contributed by species abundance of coral and organisms, nutrient supports and environmental characteristics of coral ecosystems. Coral reefs play an important ecological role of biogeochemical cycling of nutrients in waters around the reefs. These results contribute valuable information for the protection, conservation and sustainable exploitation of the natural resources in coral reef ecosystems in Vietnam.

  10. Rate determination of supercritical water gasification of primary sewage sludge as a replacement for anaerobic digestion.

    Science.gov (United States)

    Wilkinson, Nikolas; Wickramathilaka, Malithi; Hendry, Doug; Miller, Andrew; Espanani, Reza; Jacoby, William

    2012-11-01

    Supercritical water gasification of primary sewage sludge sampled from a local facility was undertaken at different solids content. The performance of the process was compared with the anaerobic digestion system in use at the facility where the samples were taken. The mass and composition of the vapor products documented showed that the process generates more energy per gram of feed while rapidly destroying more volatile solids relative to the anaerobic digestion process. However, the energy input requirements are greater for supercritical water gasification. This study defines parameters for a model of the gasification reaction using the power law and Arrhenius equation. The activation energy was estimated to be 15 kJ/mol, and the reaction order was estimated to be 0.586. This model allows estimation of the size of a supercritical water reactor needed to replace the anaerobic digesters that are currently used at the wastewater treatment plant.

  11. 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.

  12. 76 FR 33756 - Notice of Approval of the Primacy Application for National Primary Drinking Water Regulations for...

    Science.gov (United States)

    2011-06-09

    ... AGENCY Notice of Approval of the Primacy Application for National Primary Drinking Water Regulations for... following offices: (1) Missouri Department of Natural Resources, Public Drinking Water Branch, 1101... Pesticides Division, Drinking Water Management Branch, 901 North 5th Street, Kansas City, Kansas 66101....

  13. Corrosion and Corrosion Control in Light Water Reactors

    Science.gov (United States)

    Gordon, Barry M.

    2013-08-01

    Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

  14. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  15. Water use efficiency in a primary subtropical evergreen forest in Southwest China

    Science.gov (United States)

    Song, Qing-Hai; Fei, Xue-Hai; Zhang, Yi-Ping; Sha, Li-Qing; Liu, Yun-Tong; Zhou, Wen-Jun; Wu, Chuan-Sheng; Lu, Zhi-Yun; Luo, Kang; Gao, Jin-Bo; Liu, Yu-Hong

    2017-02-01

    We calculated water use efficiency (WUE) using measures of gross primary production (GPP) and evapotranspiration (ET) from five years of continuous eddy covariance measurements (2009–2013) obtained over a primary subtropical evergreen broadleaved forest in southwestern China. Annual mean WUE exhibited a decreasing trend from 2009 to 2013, varying from ~2.28 to 2.68 g C kg H2O‑1. The multiyear average WUE was 2.48 ± 0.17 (mean ± standard deviation) g C kg H2O‑1. WUE increased greatly in the driest year (2009), due to a larger decline in ET than in GPP. At the diurnal scale, WUE in the wet season reached 5.1 g C kg H2O‑1 in the early morning and 4.6 g C kg H2O‑1 in the evening. WUE in the dry season reached 3.1 g C kg H2O‑1 in the early morning and 2.7 g C kg H2O‑1 in the evening. During the leaf emergence stage, the variation of WUE could be suitably explained by water-related variables (relative humidity (RH), soil water content at 100 cm (SWC_100)), solar radiation and the green index (Sgreen). These results revealed large variation in WUE at different time scales, highlighting the importance of individual site characteristics.

  16. Lead in drinking water: sampling in primary schools and preschools in south central Kansas.

    Science.gov (United States)

    Massey, Anne R; Steele, Janet E

    2012-03-01

    Studies in Philadelphia, New York City, Houston, Washington, DC, and Greenville, North Carolina, have revealed high lead levels in drinking water. Unlike urban areas, lead levels in drinking water in suburban and rural areas have not been adequately studied. In the study described in this article, drinking water in primary schools and preschools in five suburban and rural south central Kansas towns was sampled to determine if any exceeded the U.S. Environmental Protection Agency (U.S. EPA) guidance level for schools and child care facilities of 20 parts per billion (ppb). The results showed a total of 32.1% of the samples had detectable lead levels and 3.6% exceeded the U.S. EPA guidance level for schools and child care providers of 20 ppb. These results indicate that about one-third of the drinking water consumed by children age six and under in the five suburban and rural south central Kansas towns studied has some lead contamination, exposing these children to both short-term and long-term health risks. The authors suggest a need for increased surveillance of children's drinking water in these facilities.

  17. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method; Sondas para inspecao de trocadores de calor instalados em usinas nucleares tipo PWR pelo metodo de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alonso F.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Enghenharia Mecanica]. E-mail: kauzz21@yahoo.com; Alencar, Donizete A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: daa@cdtn.br

    2007-07-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  18. 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.

  19. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    Directory of Open Access Journals (Sweden)

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  20. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  1. Numerical Simulation of Size Effects on Countercurrent Flow Limitation in PWR Hot Leg Models

    Directory of Open Access Journals (Sweden)

    I. Kinoshita

    2012-01-01

    Full Text Available We have previously done numerical simulations using the two-fluid model implemented in the CFD software FLUENT6.3.26 to investigate effects of shape of a flow channel and its size on CCFL (countercurrent flow limitation characteristics in PWR hot leg models. We confirmed that CCFL characteristics in the hot leg could be well correlated with the Wallis parameters in the diameter range of 0.05 m≤D≤0.75 m. In the present study, we did numerical simulations using the two-fluid model for the air-water tests with D=0.0254 m to determine why CCFL characteristics for D=0.0254 m were severer compared with those in the range, 0.05 m≤D≤0.75 m. The predicted CCFL characteristics agreed with the data for D=0.0254 m and indicated that the CCFL difference between D=0.0254 m and 0.05 mm≤D≤0.75 mm was caused by the size effect and not by other factors.

  2. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  3. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  4. 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.)

  5. CFD studies on the phenomena around counter-current flow limitations of gas/liquid two-phase flow in a model of a PWR hot leg

    Energy Technology Data Exchange (ETDEWEB)

    Deendarlianto, E-mail: deendarlianto@ugm.ac.id [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Department of Mechanical and Industrial Engineering, Faculty of Engineering, Gadjah Mada University, Jalan Grafika No. 2, Yogyakarta 55281 (Indonesia); Hoehne, Thomas; Lucas, Dirk; Vallee, Christophe [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Zabala, Gustavo Adolfo Montoya [Department of Chemical Engineering, Simon Bolivar University, Valle of Sartenejas, Caracas 1080 (Venezuela, Bolivarian Republic of)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We modelled CCFL in a PWR hot leg using Algebraic Interfacial Area Density model. Black-Right-Pointing-Pointer The model is able to distinguish the local flow morphologies. Black-Right-Pointing-Pointer Test fluids are air-water and steam-water. Black-Right-Pointing-Pointer Calculated CCFL and water level are in good agreement with experimental data. - Abstract: In order to improve the understanding of counter-current two-phase flow and to validate new physical models, CFD simulations of a 1/3rd scale model of the hot leg of a German Konvoi pressurized water reactor (PWR) with rectangular cross section were performed. Selected counter-current flow limitation (CCFL) experiments conducted at Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX using the multi-fluid Euler-Euler modelling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a shear stress transport (SST) turbulence model. In the simulation, the drag law was approached by a newly developed correlation of the drag coefficient in the Algebraic Interfacial Area Density (AIAD) model. The model can distinguish the bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicate also a quantitative agreement between calculations and experimental data for the CCFL characteristics and the water level inside the hot leg channel.

  6. The performance of constructed wetlands treating primary, secondary and dairy soiled water in Ireland (a review).

    Science.gov (United States)

    Healy, M G; O' Flynn, C J

    2011-10-01

    In Ireland, no database detailing the design, influent loading rates or performance of constructed wetlands (CWs) exists. On account of this, they are designed without any protocol based on empirical data. The aim of this paper was to provide the first published data on the performance of free-water surface flow (FWSF) CWs treating primary and secondary-treated municipal wastewater, and agricultural dairy soiled water (DSW) in Ireland. In total, the performance of thirty-four FWSF CWs, comprising fourteen CWs treating primary-treated municipal wastewater, thirteen CWs treating secondary-treated municipal wastewater, and seven CWs treating DSW, were examined. In most CWs, good organic, suspended solids (SS) and nutrient removal was measured. At an average organic loading rate (OLR) of 10 and 9 g biochemical oxygen demand (BOD) m(-2) d(-1), CWs treating primary and secondary wastewater removed 95 and 84% of influent BOD. Constructed wetlands treating DSW had an average BOD removal of 98%. At average SS loading rates of 6 and 14 g m(-2) d(-1), CWs treating primary and secondary wastewater had a 96 and an 82% reduction, and produced a final effluent with a concentration of 14 and 13 mg L(-1). Constructed wetlands treating DSW produced a final effluent of 34 mg L(-1) (94% reduction). Similar to other studies, all CWs examined had variable performance in ammonium-N (NH(4)(+)-N) removal, with average removals varying between 37% (for CWs treating secondary wastewater) and 88% (for CWs treating DSW). Variable ortho-phosphorus (PO(4)(3-)-P) removal was attributable to different durations of operation, media types and loading rates.

  7. 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.

  8. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  9. Mechanism for the primary transformation of acetaminophen in a soil/water system.

    Science.gov (United States)

    Liang, Chuanzhou; Lan, Zhonghui; Zhang, Xu; Liu, Yingbao

    2016-07-01

    The transformation of acetaminophen (APAP) in a soil/water system was systematically investigated by a combination of kinetic studies and a quantitative analysis of the reaction intermediates. Biotransformation was the predominant pathway for the elimination of APAP, whereas hydrolysis or other chemical transformation, and adsorption processes made almost no contribution to the transformation under a dark incubation. Bacillus aryabhattai strain 1-Sj-5-2-5-M, Klebsiella pneumoniae strain S001, and Bacillus subtilis strain HJ5 were the main bacteria identified in the biotransformation of APAP. The soil-to-water ratio and soil preincubation were able to alter the transformation kinetic pattern. Light irradiation promoted the overall transformation kinetics through enhanced biotransformation and extra photosensitized chemical reactions. The transformation pathways were strongly dependent on the initial concentration of APAP. The main primary transformation products were APAP oligomers and p-aminophenol, with the initial addition of 26.5 and 530 μM APAP, respectively. APAP oligomers accounted for more than 95% of transformed APAP, indicating that almost no bound residues were generated through the transformation of APAP in the soil/water system. The potential environmental risks of APAP could increase following the transformation of APAP in the soil/water system because of the higher toxicity of the transformation intermediates.

  10. The exchange of water between the Faroe Shelf and the surrounding waters and its effect on the primary production

    Science.gov (United States)

    Eliasen, Sólvá Karadóttir; Hansen, Bogi; Larsen, Karin Margretha Húsgarð; Hátún, Hjálmar

    2016-01-01

    The interannual variation of the spring bloom and its effect on the marine ecosystem on the Faroe Shelf has been observed for a couple of decades. However, the mechanism controlling the spring bloom has so far not been known and attempts to explain the mechanism have mostly ruled out possibilities. The Faroe Shelf is to a variable degree isolated from the surrounding waters by a tidal front. It has previously been suggested that variations in the density difference across the front and how water masses are transferred across it affect the spring primary production, which is thought to be a driver of the shelf ecosystem. Using air-sea heat flux data and sea temperature observations on the shelf and off the shelf, we estimate the cross-frontal volume exchange in January-April and find that it increases with the tidal current speed and decreases with the cross-frontal temperature difference. Using the observed exchange rates, we show that the phytoplankton growth rate may be reduced by more than 0.05 day- 1 when the exchange is intense and off-shelf production is still low. Based on frontal dynamics theory, we suggest that the cross-frontal exchange rate in the above mentioned period is determined by the rate of vertical turbulent diffusion through the front. A simple theoretical model is found to support this hypothesis qualitatively as well as quantitatively. This supports that variations in horizontal exchange are an important controlling factor of the initial spring bloom and that the horizontal exchange during the winter can be determined by vertical turbulent diffusion. Our results will be relevant for the primary production in other similar systems of small geographical extent and also for other problems involving cross-shelf exchange, such as oil spill dispersal.

  11. 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.

  12. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  13. Influence of dispersants on petroleum bioavailability to primary producers in a brackish water food chain

    Energy Technology Data Exchange (ETDEWEB)

    Younghans-Haug, C.O.; Wolfe, M.F.; Tjeerdema, R.S. [Univ. of California, Santa Cruz, CA (United States). Dept. of Chemistry and Biochemistry; Sowby, M.L. [Office of Oil Spill Prevention and Response, Sacramento, CA (United States)

    1994-12-31

    Petroleum is transported and processed within biologically rich brackish environments worldwide. Past research has investigated disposition of chemically dispersed oil in mammals, fish, and higher invertebrates, yet little is known about how chemical dispersion influences petroleum behavior within primary producers within brackish water food chains. One concern is whether chemical dispersion influences petroleum bioavailability to primary producers and the potential for increased petroleum bioaccumulation. This research examines changes in petroleum bioavailability to the euryhaline phytoplankton Isochrysis galbana by measuring bioconcentration factors (BCFS) including uptake and deputation rates for dispersed and undispersed brackish water oil spills. Isochrysis is a major food source for zooplankton which are consumed by a multitude of larval fish having both ecological and commercial importance. Prudhoe Bay Crude oil, Corexit 9527, and {sup 14}C-naphthalene were used for these studies. Constant-concentration flow-through exposures were employed for the uptake and BCF experiments. Work was performed below the ``no observable effect concentration`` to eliminate stress-induced metabolic altercations that could in themselves influence petroleum behavior. Exposure chamber and experimental design will be discussed, and study results presented. Understanding how chemical dispersion alters petroleum behavior within the lowest levels of the food chain leads to better delineation of consumer risks.

  14. Gas-liquid countercurrent two-phase flow in a PWR hot leg: A comprehensive research review

    Energy Technology Data Exchange (ETDEWEB)

    Deendarlianto, E-mail: deendarlianto@ugm.ac.id [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Department of Mechanical and Industrial Engineering, Faculty of Engineering, Gadjah Mada University, Jalan Grafika No. 2, Yogyakarta 55281 (Indonesia); Hoehne, Thomas; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Vierow, Karen [Department of Nuclear Engineering Texas A and M University, 129 Zachry Engineering Center, 3133 TAMU College Station, TX 77843-3133 (United States)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer We review the scientific progress on the CCFL in a PWR hot leg. Black-Right-Pointing-Pointer It includes the experimental data, one-dimensional and CFD models in the open literatures. Black-Right-Pointing-Pointer The weak and strong points of the published works were clarified. Black-Right-Pointing-Pointer The research directions in this field were proposed. - Abstract: Research into gas-liquid countercurrent two-phase flow in a model of pressurized water reactor (PWR) hot leg has been carried out over the last several decades. An extensive experimental data base has been accumulated from these studies, leading to the development of phenomenological correlations and scaling parameters of the countercurrent flow limitation (CCFL). However, most of the proposed correlations apply under a relatively narrow range of conditions, generally limited to the test section conditions and/or geometry. Moreover the development of mechanistic models based on the underlying physical processes has been limited. In contrast to this mechanistic form of modelling, the implementation of computational fluid dynamics (CFD) techniques has also been pursued, but the considerable robust three-dimensional (3D) closure relations for this application remain an unachieved goal due to lack of detailed phenomenological knowledge and consequent application of empirical one-dimensional experimental correlations to the multidimensional problem. This paper presents a comprehensive review of research work on countercurrent gas-liquid two-phase flow in a PWR hot leg and provides direction regarding future research on this topic. In the introductory section, the problems facing current research are described. In the following sections, recent experimental as well as theoretical research achievements are overviewed. In the last section, the problems that remain unsolved are discussed, along with some concluding remarks. It was found that only limited theoretical

  15. Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

    Energy Technology Data Exchange (ETDEWEB)

    Han, Guangdong [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, Shanghai, 200072 (China); Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Shanghai, 200072 (China)

    2015-12-15

    The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen. - Highlights: • Duplex oxide films on ground and mechanically polished specimens. • Compact oxide on electropolished specimen after 120 h immersion. • Large spinel outer layer rich in Fe and fine spinel inner layer rich in Cr. • Electropolishing improved oxidation resistance especially at the early stages. • Inhomogeneous Cr-rich inner layer with granular areas affected by surface treatment.

  16. Screening and Quantification of Aliphatic Primary Alkyl Corrosion Inhibitor Amines in Water Samples by Paper Spray Mass Spectrometry.

    Science.gov (United States)

    Jjunju, Fred P M; Maher, Simon; Damon, Deidre E; Barrett, Richard M; Syed, S U; Heeren, Ron M A; Taylor, Stephen; Badu-Tawiah, Abraham K

    2016-01-19

    Direct analysis and identification of long chain aliphatic primary diamine Duomeen O (n-oleyl-1,3-diaminopropane), corrosion inhibitor in raw water samples taken from a large medium pressure water tube boiler plant water samples at low LODs (corrosion inhibitors in an industrial water boiler plant and other related samples in the water treatment industry. This approach was applied for the analysis of three complex water samples including feedwater, condensate water, and boiler water, all collected from large medium pressure (MP) water tube boiler plants, known to be dosed with varying amounts of polyamine and amine corrosion inhibitor components. Polyamine chemistry is widely used for example in large high pressure (HP) boilers operating in municipal waste and recycling facilities to prevent corrosion of metals. The samples used in this study are from such a facility in Coventry waste treatment facility, U.K., which has 3 × 40 tonne/hour boilers operating at 17.5 bar.

  17. 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.

  18. 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.

  19. 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.

  20. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  1. 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.

  2. Global gross primary productivity and water use efficiency changes under drought stress

    Science.gov (United States)

    Yu, Zhen; Wang, Jingxin; Liu, Shirong; Rentch, James S.; Sun, Pengsen; Lu, Chaoqun

    2017-01-01

    Drought can affect the structure, composition and function of terrestrial ecosystems, yet drought impacts and post-drought recovery potentials of different land cover types have not been extensively studied at a global scale. We evaluated drought impacts on gross primary productivity (GPP), evapotranspiration (ET), and water use efficiency (WUE) of different global terrestrial ecosystems, as well as the drought-resilience of each ecosystem type during the period of 2000 to 2011. Using GPP as biome vitality indicator against drought stress, we developed a model to examine ecosystem resilience represented by the length of recovery days (LRD). LRD presented an evident gradient of high (>60 days) in mid-latitude region and low (stress induced responses or data uncertainties, which require further investigation.

  3. 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.

  4. 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)

  5. 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.

  6. Factors associated with use of improved water sources and sanitation among rural primary schoolchildren in Pursat Province, Cambodia.

    Science.gov (United States)

    En, Wee Liang; Gan, Goh Lee

    2011-07-01

    Access to improved water supplies and sanitation generally reduces childhood diarrhea incidence. Using a cross-sectional stratified cluster sampling design, interviews were conducted among grade 4-6 primary schoolchildren from 10 primary schools in highland and lowland districts of Pursat Province, Cambodia, in both June (rainy season) and December (dry season) 2009 to determine the demographics and water sources/sanitation used. Parents also recorded any incidents of diarrhea in their children over those months. We explored the sociodemographic factors associated with use of improved water sources/sanitation, using mixed effect modelling. Participation was 84.7% (1,101/1,300). About half exclusively used improved water sources but less than 25% had access to improved sanitation during both seasons. Adjusting for clustering within households and within individuals over time, exclusive use of improved water sources and sanitation were associated with the following: dry season, more permanent housing type, family size improved sanitation was associated with good hygiene practices and exclusive use of improved water sources was associated with male gender. Access to improved water sources and sanitation among rural Cambodian primary schoolchildren can be improved, particularly in those with lower socio-economic status. Programs to promote use of improved water sources/sanitation need to target less educated parents.

  7. 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

  8. 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.

  9. Factors leading to poor water sanitation hygiene among primary school going children in Chitungwiza

    Directory of Open Access Journals (Sweden)

    Blessing Dube

    2012-03-01

    Full Text Available Although the world has progressed in the area of water and sanitation, more than 2.3 billion people still live without access to sanitation facilities and some are unable to practice basic hygiene. Access to water and basic sanitation has deteriorated in Chitungwiza and children are at risk of developing illness and missing school due to the deterioration. We sought to investigate the predisposing, enabling and reinforcing factors that are causally related to water- and sanitation- related hygiene practices among school going children. A random sample of 400 primary school children (196 males, 204 females in four schools in Chitungwiza town, Zimbabwe was interviewed. Behavioural factors were assessed through cross examination of the PROCEED PRECEDE Model. The respondents had been stratified through the random sampling where strata were classes. A structured observation checklist was also administered to assess hygiene enabling facilities for each school. Children’s knowledge and perceptions were inconsistent with hygienic behaviour. The family institution seemed to play a more important role in life skills training and positive reinforcement compared to the school (50% vs 27.3%. There was no association between a child’s sex, age and parents’ occupation with any of the factors assessed (P=0.646. Schools did not provide a hygiene enabling environment as there were no learning materials, policy and resources on hygiene and health. The challenges lay in the provision of hygiene enabling facilities, particularly, the lack of access to sanitation for the maturing girl child and a school curriculum that provides positive reinforcement and practical life skills training approach.

  10. Numerical investigation on stress corrosion cracking behavior of dissimilar weld joints in pressurized water reactor plants

    Directory of Open Access Journals (Sweden)

    Lingyan Zhao

    2014-07-01

    Full Text Available There have been incidents recently where stress corrosion cracking (SCC observed in the dissimilar metal weld (DMW joints connecting the reactor pressure vessel (RPV nozzle with the hot leg pipe. Due to the complex microstructure and mechanical heterogeneity in the weld region, dissimilar metal weld joints are more susceptible to SCC than the bulk steels in the simulated high temperature water environment of pressurized water reactor (PWR. Tensile residual stress (RS, in addition to operating loads, has a great contribution to SCC crack growth. Limited experimental conditions, varied influence factors and diverging experimental data make it difficult to accurately predict the SCC behavior of DMW joints with complex geometry, material configuration, operating loads and crack shape. Based on the film slip/dissolution oxidation model and elastic-plastic finite element method (EPFEM, an approach is developed to quantitatively predict the SCC growth rate of a RPV outlet nozzle DMW joint. Moreover, this approach is expected to be a pre-analytical tool for SCC experiment of DMW joints in PWR primary water environment.

  11. Water-polysaccharide interactions in the primary cell wall of Arabidopsis thaliana from polarization transfer solid-state NMR.

    Science.gov (United States)

    White, Paul B; Wang, Tuo; Park, Yong Bum; Cosgrove, Daniel J; Hong, Mei

    2014-07-23

    Polysaccharide-rich plant cell walls are hydrated under functional conditions, but the molecular interactions between water and polysaccharides in the wall have not been investigated. In this work, we employ polarization transfer solid-state NMR techniques to study the hydration of primary-wall polysaccharides of the model plant, Arabidopsis thaliana. By transferring water (1)H polarization to polysaccharides through distance- and mobility-dependent (1)H-(1)H dipolar couplings and detecting it through polysaccharide (13)C signals, we obtain information about water proximity to cellulose, hemicellulose, and pectins as well as water mobility. Both intact and partially extracted cell wall samples are studied. Our results show that water-pectin polarization transfer is much faster than water-cellulose polarization transfer in all samples, but the extent of extraction has a profound impact on the water-polysaccharide spin diffusion. Removal of calcium ions and the consequent extraction of homogalacturonan (HG) significantly slowed down spin diffusion, while further extraction of matrix polysaccharides restored the spin diffusion rate. These trends are observed in cell walls with similar water content, thus they reflect inherent differences in the mobility and spatial distribution of water. Combined with quantitative analysis of the polysaccharide contents, our results indicate that calcium ions and HG gelation increase the amount of bound water, which facilitates spin diffusion, while calcium removal disrupts the gel and gives rise to highly dynamic water, which slows down spin diffusion. The recovery of spin diffusion rates after more extensive extraction is attributed to increased water-exposed surface areas of the polysaccharides. Water-pectin spin diffusion precedes water-cellulose spin diffusion, lending support to the single-network model of plant primary walls in which a substantial fraction of the cellulose surface is surrounded by pectins.

  12. Understanding of Hydriding Mechanisms of Zircaloy-4 Alloy during Corrosion in PWR Simulated Conditions and Influence of Zirconium Hydrides on Zircaloy-4 Corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bisor-Melloul, C.; Tupin, M.; Bossis, P.; Chene, J.; Bechade, J.L. [CEA Saclay, 91 - Gif sur Yvette (France); Motta, A. [Pennsylvania State Univ. (United States)

    2011-03-15

    Zirconium alloys represent the first containment barrier to fission products, their mechanical integrity is essential for nuclear safety in PWR. During their corrosion in primary water, some of the hydrogen involved in the oxidation reaction with water ingresses into the alloy through the oxide layer. In the metallic matrix, once the solid solution limit is reached at the irradiation temperature, hydrogen precipitates as Zr hydrides mainly located just under the metal/oxide interface due to the thermal gradient across the cladding. As these hydrides may contribute to a larger oxide thickness and to a more fragile behaviour of the cladding, the minimization of hydrogen pick-up is required. Accordingly, since the Zircaloy-4 (Zr-1.3Sn-0.2Fe-0.1Cr) alloy is known to be sensitive to this phenomenon, the understanding of its hydriding mechanism, isotopic exchanges were carried out in D{sub 2}O environment at 360 C and led to the localization, in the oxide scales, of the limiting step for the hydrogen diffusion. To estimate an apparent diffusion coefficient of hydrogen in the oxide formed on Zircaloy-4, we based on SIMS profiles and penetration depth of deuterium in the dense part of the oxide film. Then ERDA estimation of the hydrogen content in zirconia and fusion measurement of the hydrogen content in both metal and oxide were used to estimate a hydrogen flux absorbed by the alloy and hence to deduce an apparent diffusion coefficient. Finally, these 2 methods lead to quite similar values (between 1.10{sup -14} cm{sup 2}/s and 6.10{sup -14} cm{sup 2}/s) which are in accordance with bibliography. Concerning the impact of hydrides on the corrosion of Zircaloy-4, several pre-hydrided and reference samples were corroded simultaneously at 360 C. The characterization of the pre-hydrided samples revealed some changes, as the presence of the Zr{sub 3}O sub-oxide at the inner metal/oxide interface, a lower fraction of -ZrO{sub 2} in the oxide and a faster diffusion of oxygen

  13. Seasonal Shifts in Primary Water Source Type: A Comparison of Largely Pastoral Communities in Uganda and Tanzania.

    Science.gov (United States)

    Pearson, Amber L; Zwickle, Adam; Namanya, Judith; Rzotkiewicz, Amanda; Mwita, Emiliana

    2016-01-27

    Many water-related illnesses show an increase during the wet season. This is often due to fecal contamination from runoff, yet, it is unknown whether seasonal changes in water availability may also play a role in increased illness via changes in the type of primary water source used by households. Very little is known about the dynamic aspects of access to water and changes in source type across seasons, particularly in semi-arid regions with annual water scarcity. The research questions in this study were: (1) To what degree do households in Uganda (UG) and Tanzania (TZ) change primary water source type between wet and dry seasons?; and (2) How might seasonal changes relate to water quality and health? Using spatial survey data from 92 households each in UG and TZ this study found that, from wet to dry season, 26% (UG) and 9% (TZ) of households switched from a source with higher risk of contamination to a source with lower risk. By comparison, only 20% (UG) and 0% (TZ) of households switched from a source with lower risk of contamination to a source with higher risk of contamination. This research suggests that one pathway through which water-related disease prevalence may differ across seasons is the use of water sources with higher risk contamination, and that households with access to sources with lower risks of contamination sometimes choose to use more contaminated sources.

  14. Synthesis of primary production in the Arctic Ocean: I. Surface waters, 1954-2007

    Science.gov (United States)

    Matrai, P. A.; Olson, E.; Suttles, S.; Hill, V.; Codispoti, L. A.; Light, B.; Steele, M.

    2013-03-01

    The spatial and seasonal magnitude and variability of primary production in the Arctic Ocean (AO) is quantified with a pan-arctic approach. We synthesize estimates of primary production (PP), focusing on surface waters (0-5 m), using complementary methods that emphasize different spatial and temporal scales. These methods include (1) in situ observations of 14C uptake mostly and possibly some O2 production reported in units of carbon (in situ PP), (2) remotely sensed primary production (sat-PP), and (3) an empirical algorithm giving net PP as a function of in situ chlorophyll a (in situ Chl-PP). The work presented herein examines historical data for PP collected in surface waters only, as they form the majority of the values of a larger ensemble of PP data collected over >50 years (ARCSS-PP) by many national and international efforts. This extended set of surface and vertically-resolved data will provide pan-Arctic validation of remotely sensed chlorophyll a and PP, an extremely valuable tool in this environment which is so difficult to sample. To this day, PP data in the AO are scarce and have uneven temporal and spatial coverage which, when added to the AO’s regional heterogeneity, its strong seasonal changes, and limited access, have made and continue to make obtaining a comprehensive picture of PP in the AO difficult. Daily surface in situ PP averaged 70 and 21 mg C m-3 d-1 for spring and summer, respectively, for the ca. 50 year period across the AO. Average daily estimates of in situ PP in surface waters on a pan-Arctic basis were several fold higher with respect to remotely sensed PP (sat-PP) and in situ chlorophyll-derived PP (Chl-PP) in the spring period, likely due to differences in data availability and coverage. Summer daily averages for surface in situ PP and sat-PP were similar and twice as high as in situ Chl-PP. Differences among annual estimates of surface in situ PP, in situ Chl-PP and sat-PP across the Arctic Ocean are presented and discussed

  15. Effect of supplementation of tender coconut water on blood pressure of primary hypertensive subjects

    Directory of Open Access Journals (Sweden)

    Gullapalli HS, Avinash P Tekade, Namrata H Gullapalli

    2013-04-01

    Full Text Available Background: Hypertension is a major health problem worldwide. Increased vascular resistance, sodium retention & sympathetic over activity contributes to the blood pressure elevation. Plant foods may be beneficial in decreasing blood pressure (BP. Recently much attention has been focused on plant foods that may be beneficial in preventing Hypertension, metabolic syndrome and possibly reduce the risk of various diseases. This clinical study was conducted to test the effectiveness of a structured intervention on BP of primary hypertensive subjects. Aim: To study the effect of Tender Coconut Water (TCW on BP of Primary hypertensive subjects. Methods and Material: 70 subjects were selected randomly sample for 6 weeks of the intervention program. Among them 40 subjects were selected as the experimental group and 30 300ml/day for 6 weeks whereas the control group was instructed to follow the same routine without modifications. One initial, two mid intervention (after every 15 days and one final (post intervention BP recorded for both the groups. The obtained data was statistically analyzed. Results: The mean systolic BP of experimental group and control group were decreased from 145.8 mm Hg and 141mm of Hg to 135.3 mm of Hg and 140 mm of Hg respectively. The mean diastolic BP of experimental group and control group were decreased from 93.7 mm H g and 90.9 mmHg to 86.9 mm of Hg and 89.7 mm of Hg respectively. Conclusion: Irrespective of cause of hypertension TCW has beneficial effect on BP. TCW contains high amount of potassium which causes vasodilatation and also improve the endothelial function.

  16. 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

  17. Calculation of water-bearing primary basalt and estimation of source mantle conditions beneath arcs: PRIMACALC2 model for WINDOWS

    Science.gov (United States)

    Kimura, Jun-Ichi; Ariskin, Alexey A.

    2014-04-01

    present a new method for estimating the composition of water-bearing primary arc basalt and its source mantle conditions. The PRIMACALC2 model uses a thermodynamic fractional crystallization model COMAGMAT3.72 and runs with an Excel macro to examine the mantle equilibrium and trace element calculations of a primary basalt. COMAGMAT3.72 calculates magma fractionation in 0-10 kb at various compositions, pressure, oxygen fugacity, and water content, but is only applicable for forward calculations. PRIMACALC2 first calculates the provisional composition of a primary basalt from an observed magma. The basalt composition is then calculated by COMAGMAT3.72 for crystallization. Differences in elemental concentrations between observed and the closest-match calculated magmas are then adjusted in the primary basalt. Further iteration continues until the calculated magma composition converges with the observed magma, resulting in the primary basalt composition. Once the fitting is satisfied, back calculations of trace elements are made using stepwise addition of fractionated minerals. Mantle equilibrium of the primary basalt is tested using the Fo-NiO relationship of olivine in equilibrium with the primary basalt, and thus with the source mantle. Source mantle pressure, temperature, and degree of melting are estimated using petrogenetic grids based on experimental data obtained in anhydrous systems. Mantle melting temperature in a hydrous system is computed by adjusting T with a parameterization for a water-bearing system. PRIMACALC2 can be used either in dry or water-bearing arc magmas and is also applicable to mid-ocean ridge basalts and nonalkalic ocean island basalts.

  18. Plasma membrane proteomics in the maize primary root growth zone: novel insights into root growth adaptation to water stress.

    Science.gov (United States)

    Voothuluru, Priyamvada; Anderson, Jeffrey C; Sharp, Robert E; Peck, Scott C

    2016-09-01

    Previous work on maize (Zea mays L.) primary root growth under water stress showed that cell elongation is maintained in the apical region of the growth zone but progressively inhibited further from the apex. These responses involve spatially differential and coordinated regulation of osmotic adjustment, modification of cell wall extensibility, and other cellular growth processes that are required for root growth under water-stressed conditions. As the interface between the cytoplasm and the apoplast (including the cell wall), the plasma membrane likely plays critical roles in these responses. Using a simplified method for enrichment of plasma membrane proteins, the developmental distribution of plasma membrane proteins was analysed in the growth zone of well-watered and water-stressed maize primary roots. The results identified 432 proteins with differential abundances in well-watered and water-stressed roots. The majority of changes involved region-specific patterns of response, and the identities of the water stress-responsive proteins suggest involvement in diverse biological processes including modification of sugar and nutrient transport, ion homeostasis, lipid metabolism, and cell wall composition. Integration of the distinct, region-specific plasma membrane protein abundance patterns with results from previous physiological, transcriptomic and cell wall proteomic studies reveals novel insights into root growth adaptation to water stress.

  19. Mechanical factors in primary water stress corrosion cracking of cold-worked stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Hammadi, Rashid Al, E-mail: rashid.alhammadi@fanr.gov.ae [Nuclear Security Division, Federal Authority for Nuclear Regulation, Abu Dhabi (United Arab Emirates); Yi, Yongsun, E-mail: yongsun.yi@kustar.ac.ae [Department of Nuclear Engineering, Khalifa University, Abu Dhabi (United Arab Emirates); Zaki, Wael, E-mail: wael.zaki@kustar.ac.ae [Department of Mechanical Engineering, Khalifa University, Abu Dhabi (United Arab Emirates); Cho, Pyungyeon, E-mail: pyungyeon.cho@kustar.ac.ae [Department of Nuclear Engineering, Khalifa University, Abu Dhabi (United Arab Emirates); Jang, Changheui, E-mail: chjang@kaist.ac.kr [Nuclear and Quantum Engineering Department, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2016-05-15

    Highlights: • PWSCC of cold-worked austenitic stainless steel was studied. • Finite element analysis was performed on a compact tension specimen. • Mechanical fields near a crack tip were evaluated using FEA. • The dependence of mechanical factors on K{sub I} and yield stress was investigated. • The crack tip normal stress was identified as a main factor controlling PWSCC. - Abstract: Finite element analysis was performed on a compact tension specimen to determine the stress and strain distributions near a crack tip. Based on the results, the crack tip stain rates by crack advance and creep rates near crack tip were estimated. By comparing the dependence of the mechanical factors on the stress intensity factor and yield stress with that of the SCC crack growth rates, it was tried to identify the main mechanical factor for the primary water stress corrosion cracking (PWSCC) of cold-worked austenitic stainless steels. The analysis results showed that the crack tip normal stress could be the main mechanical factor controlling the PWSCC, suggesting that the internal oxidation mechanism might be the most probable PWSCC mechanism of cold-worked austenitic stainless steels.

  20. The Life-Cycle Costs of School Water, Sanitation and Hygiene Access in Kenyan Primary Schools

    Directory of Open Access Journals (Sweden)

    Kelly T. Alexander

    2016-06-01

    Full Text Available Water, Sanitation and Hygiene (WASH programs in schools can increase the health, dignity and comfort of students and teachers. Understanding the costs of WASH facilities and services in schools is one essential piece for policy makers to utilize when budgeting for schools and helping to make WASH programs more sustainable. In this study we collected data from NGO and government offices, local hardware shops and 89 rural primary schools across three Kenyan counties. Current expenditures on WASH, from school and external (NGO, government, parent sources, averaged 1.83 USD per student per year. After reviewing current expenditures, estimated costs of operations and maintenance for bringing schools up to basic WASH standards, were calculated to be 3.03 USD per student per year. This includes recurrent costs, but not the cost of installing or setting up WASH infrastructure, which was 18,916 USD per school, for a school of 400 students (4.92 USD per student, per year. These findings demonstrate the need for increases in allocations to schools in Kenya, and stricter guidance on how money should be spent on WASH inputs to enable all schools to provide basic WASH for all students.

  1. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  2. The Life-Cycle Costs of School Water, Sanitation and Hygiene Access in Kenyan Primary Schools.

    Science.gov (United States)

    Alexander, Kelly T; Mwaki, Alex; Adhiambo, Dorothy; Cheney-Coker, Malaika; Muga, Richard; Freeman, Matthew C

    2016-06-27

    Water, Sanitation and Hygiene (WASH) programs in schools can increase the health, dignity and comfort of students and teachers. Understanding the costs of WASH facilities and services in schools is one essential piece for policy makers to utilize when budgeting for schools and helping to make WASH programs more sustainable. In this study we collected data from NGO and government offices, local hardware shops and 89 rural primary schools across three Kenyan counties. Current expenditures on WASH, from school and external (NGO, government, parent) sources, averaged 1.83 USD per student per year. After reviewing current expenditures, estimated costs of operations and maintenance for bringing schools up to basic WASH standards, were calculated to be 3.03 USD per student per year. This includes recurrent costs, but not the cost of installing or setting up WASH infrastructure, which was 18,916 USD per school, for a school of 400 students (4.92 USD per student, per year). These findings demonstrate the need for increases in allocations to schools in Kenya, and stricter guidance on how money should be spent on WASH inputs to enable all schools to provide basic WASH for all students.

  3. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Science.gov (United States)

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  4. Water use efficiency of net primary production in global terrestrial ecosystems

    Indian Academy of Sciences (India)

    Lei Xia; Fei Wang; Xingmin Mu; Kai Jin; Wenyi Sun; Peng Gao; Guangju Zhao

    2015-07-01

    The carbon and water cycles of terrestrial ecosystems, which are strongly coupled via water use efficiency (WUE), are influenced by global climate change. To explore the relationship between the carbon and water cycles and predict the effect of climate change on terrestrial ecosystems, it is necessary to study the WUE in global terrestrial ecosystems. In this study, the 13-year WUE (i.e., net primary production (NPP)/evapotranspiration (ET)) of global terrestrial ecosystems was calculated based on the Moderate Resolution Imaging Spectro-radiometer (MODIS) NPP (MOD17A3) and ET (MOD16A3) products from 2000 to 2012. The results indicate that the annual average WUE decreased but not significantly, and the 13-year mean value was 868.88 mg C m−2 mm−1. The variation trend of WUE value for each pixel differed greatly across the terrestrial ecosystems. A significant variation ( < 0.05) occurred in about 18.50% of the land surface. WUE was spatially distributed from 0 to 2541 mg C m−2 mm−1, and 58.78% of the WUE values were concentrated in the interval of 600–1200 mg C m−2 mm−1. The WUE increased from north to south in Africa and Oceania and from east to west in Europe and South America. Both latitudinal and longitudinal gradients existed in Asia and North America. The following trends in the WUE of different continents and Köppen–Geiger climates were observed: Europe (1129.71 mg C m−2 mm−1) > Oceania (1084.46 mg C m−2 mm−1) > Africa (893.51 mg C m−2 mm−1) > South America (893.07 mg C m−2 mm−1) > North America (870.79 mg C m−2 mm−1) > Asia (738.98 mg C m−2 mm−1) and warm temperate climates (1094 mg C m−2 mm−1) > snowy climates (862 mg C m−2 mm−1) > arid climates (785 mg C m−2 mm−1) > equatorial climates (732 mg C m−2 mm−1) > polar climates (435 mg C m−2 mm−1). Based on the WUE value and the present or future rainfall, the maximum carbon that fixed in one region may be theoretically calculated. Also, under the

  5. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  6. 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.

  7. The first association of a primary amebic meningoencephalitis death with culturable Naegleria fowleri in tap water from a US treated public drinking water system.

    Science.gov (United States)

    Cope, Jennifer R; Ratard, Raoult C; Hill, Vincent R; Sokol, Theresa; Causey, Jonathan Jake; Yoder, Jonathan S; Mirani, Gayatri; Mull, Bonnie; Mukerjee, Kimberly A; Narayanan, Jothikumar; Doucet, Meggie; Qvarnstrom, Yvonne; Poole, Charla N; Akingbola, Olugbenga A; Ritter, Jana M; Xiong, Zhenggang; da Silva, Alexandre J; Roellig, Dawn; Van Dyke, Russell B; Stern, Harlan; Xiao, Lihua; Beach, Michael J

    2015-04-15

    Naegleria fowleri is a climate-sensitive, thermophilic ameba found in warm, freshwater lakes and rivers. Primary amebic meningoencephalitis (PAM), which is almost universally fatal, occurs when N. fowleri-containing water enters the nose, typically during swimming, and migrates to the brain via the olfactory nerve. In August 2013, a 4-year-old boy died of meningoencephalitis of unknown etiology in a Louisiana hospital. Clinical and environmental testing and a case investigation were initiated to determine the cause of death and to identify potential exposures. Based on testing of cerebrospinal fluid and brain specimens, the child was diagnosed with PAM. His only reported water exposure was tap water; in particular, tap water that was used to supply water to a lawn water slide on which the child had played extensively prior to becoming ill. Water samples were collected from both the home and the water distribution system that supplied the home and tested; N. fowleri was identified in water samples from both the home and the water distribution system. This case is the first reported PAM death associated with culturable N. fowleri in tap water from a US treated drinking water system. This case occurred in the context of an expanding geographic range for PAM beyond southern states, with recent case reports from Minnesota, Kansas, and Indiana. This case also highlights the role of adequate disinfection throughout drinking water distribution systems and the importance of maintaining vigilance when operating drinking water systems using source waters with elevated temperatures. Published by Oxford University Press on behalf of the Infectious Diseases Society of America 2015. This work is written by (a) US Government employee(s) and is in the public domain in the US.

  8. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  9. 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.

  10. Extraction Tools for Identification of Chemical Contaminants in Estuarine and Coastal Waters to Determine Toxic Pressure on Primary Producers

    NARCIS (Netherlands)

    Booij, P; Sjollema, S.B.; Leonards, P.E.G.; de Voogt, P.; Stroomberg, G.J.; Vethaak, A.D.; Lamoree, M.H.

    2013-01-01

    The extent to which chemical stressors affect primary producers in estuarine and coastal waters is largely unknown. However, given the large number of legacy pollutants and chemicals of emerging concern present in the environment, this is an important and relevant issue that requires further study.

  11. Absorption-based algorithm of primary production for total and size-fractionated phytoplankton in coastal waters

    NARCIS (Netherlands)

    Barnes, M.K.; Tilstone, G.H.; Smyth, T.J.; Suggett, D.J.; Astoreca, R.; Lancelot, C.; Kromkamp, J.C.

    2014-01-01

    Most satellite models of production have been designed and calibrated for use in the open ocean. Coastal waters are optically more complex, and the use of chlorophyll a (chl a) as a first-order predictor of primary production may lead to substantial errors due to significant quantities

  12. Bidirectional reflectance spectroscopy of carbonaceous chondrites: Implications for water quantification and primary composition

    Science.gov (United States)

    Garenne, A.; Beck, P.; Montes-Hernandez, G.; Brissaud, O.; Schmitt, B.; Quirico, E.; Bonal, L.; Beck, C.; Howard, K. T.

    2016-01-01

    In this study, we measured bidirectional reflectance spectra (0.5-4.0 μm) of 24 CMs, five CRs, one CI, one CV, and one C2 carbonaceous chondrites. These meteorites are known to have experienced an important variability in their relative degrees of aqueous alteration degree (Rubin et al. [2007]. Geochim. Cosmochim. Acta 71, 2361-2382; Howard et al. [2009]. Geochim. Cosmochim. Acta 73, 4576-4589; Howard et al. [2011]. Geochim. Cosmochim. Acta 75, 2735-2751; Alexander et al. [2013]. Geochim. Cosmochim. Acta 123, 244-260). These measurements were performed on meteorite powders inside an environmental cell under a primary vacuum and heated at 60 °C in order to minimize adsorbed terrestrial water. This protocol allows controlling of atmospheric conditions (i.e. humidity) in order to avoid contamination by terrestrial water. We discuss various spectral metrics (e.g. reflectance, band depth, single-scattering albedo, …) in the light of recent bulk composition characterization (Howard et al. [2009]. Geochim. Cosmochim. Acta 73, 4576-4589; Howard et al. [2015]. Geochim. Cosmochim. Acta 149, 206-222; Alexander et al. [2012]. Science 337, 721; Beck et al. [2014]. Icarus 229, 263-277; Garenne et al. [2014]. Geochim. Cosmochim. Acta 137, 93-112). This study reveals variability of reflectance among meteorite groups. The reflectance is not correlated with carbon or hydrogen abundance neither with measured grain size distribution. We suggest that it is rather controlled by the nature of accreted components, in particular the initial matrix/chondrule proportion. Band depth, integrated band depth, mean optical path length, normalized optical path length, effective single-particle absorption thickness were calculated on the so called 3-μm band for reflectance spectra and for single scattering albedo spectra. They were compared with hydrated phase proportions from previous study on the same meteorites by thermogravimetric analyses and infrared spectroscopy in transmission. We find

  13. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  14. The k-[epsilon] modeling of deboration transients in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Oosterkamp, W.J.; Termaat, K.P.; Verhagen, F.C.M. (N.V. KEMA, Arnhem (Netherlands))

    1992-01-01

    The potential for reactivity accidents is receiving more attention after the Chernobyl disaster. Boron dilution transients are one class of reactivity accidents possible in pressurized water reactors (PWRs). Severe boron dilution reactivity accidents can only occur when three conditions are met: (1) a source of nonborated water is attaached to the primary system; (2) conditions are such that this nonborated water accumulates undetected outside the core; and (3) the nonborated water is rapidly moved into the core.

  15. Investigation into the effect of water chemistry on corrosion product formation in areas of accelerated flow

    Science.gov (United States)

    McGrady, John; Scenini, Fabio; Duff, Jonathan; Stevens, Nicholas; Cassineri, Stefano; Curioni, Michele; Banks, Andrew

    2017-09-01

    The deposition of CRUD (Chalk River Unidentified Deposit) in the primary circuit of a Pressurised Water Reactor (PWR) is known to preferentially occur in regions of the circuit where flow acceleration of coolant occurs. A micro-fluidic flow cell was used to recreate accelerated flow under simulated PWR conditions, by flowing water through a disc with a central micro-orifice. CRUD deposition was reproduced on the disc, and CRUD Build-Up Rates (BUR) in various regions of the disc were analysed. The effect of the local environment on BUR was investigated. In particular, the effect of flow velocity, specimen material and Fe concentration were considered. The morphology and composition of the deposits were analysed with respect to experimental conditions. The BUR of CRUD was found to be sensitive to flow velocity and Fe concentration, suggesting that mass transfer is an important factor. The morphology of the deposit was affected by the specimen material indicating a dependence on surface/particle electrostatics meaning surface chemistry plays an important role in deposition. The preferential deposition of CRUD in accelerated flow regions due to electrokinetic effects was observed and it was shown that higher Fe concentrations in solution increased BURs within the orifice whereas increased flow velocity reduced BURs.

  16. 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

  17. Example Calculations of In{sub v}essel Steam Explosions for a Prototypical PWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ik Kyu; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In this paper, the sample calculation for the in{sub v}essel steam explosions were done by using the MC3D code. The evaluation of the computational code had been done against TROI experiments and the code had been adapted to a PWR ex{sub v}essel steam explosion calculations. MC3D is a code for the calculation of different types of multiphase multi-component flows. It has been built with the fuel-coolant interaction calculations in mind. It is, however, able to calculate very different situations and has a rather wide field of potential applications. MC3D is a set of two fuel-coolant interaction codes with a common numeric solver, one for the premixing phase and one for the explosion phase. In general, the steam explosion simulation with MC3D is being carried out in two steps. In the first step, the distributions of the melt, water, and vapor phases at steam explosion triggering are being calculated with the premixing module. These premixing simulation results present the input for the second step when the escalation and propagation of the steam explosion through the premixture are being calculated with the explosion module. The MC3D premixing model is a six-field application in which the melt is described by three fields. The first one is called 'continuous' and can describe many situations as, e.g., a jet or the melt lying on the bottom of a vessel. The second field corresponds to the droplets issued from the jet fragmentation. This field describes the discontinuous state of the fuel. The third field is optional and describes the fuel fragments issuing from drop fine fragmentation. The remaining three fields are the water, the vapor, and a noncondensable gas. The drop surface area is modeled with a standard interfacial area transport equation. In the explosion model, the continuous phase is not present and only the two fields related to the dispersed fuel are considered

  18. Nickel Alloy Primary Water Bulk Surface and SCC Corrosion Film Analytical Characterization and SCC Mechanistic Implications

    Energy Technology Data Exchange (ETDEWEB)

    Morton, D.; Lewis, N.; Hanson, M.; Rice, S.; Sanders, P.

    2007-04-18

    Alloy 600 corrosion coupon tests were performed: (1) to quantify the temperature dependency of general corrosion and (2) to characterize the composition and structure of bulk surface corrosion films for comparison with ongoing primary water SCC (PWSCC) crack tip corrosion film analyses. Results suggest that the thermal activation energy of Alloy 600 corrosion is consistent with the thermal activation energy of nickel alloy PWSCC. Analytical investigations of the structure and composition of Alloy 600 bulk surface corrosion oxides revealed a duplex (inner and outer) oxide layer structure. The outer layer is discontinuous and comprised of relatively large (1 to 3 {micro}m) nickel ferrite crystals and smaller ({approx}0.1 {micro}m) chromium containing nickel ferrite crystals. The inner layer consists of a relatively continuous chromite spinel (major phase) and chromia (Cr{sub 2}O{sub 3} minor phase) which formed through non-selective oxidation. Chromia and dealloyed Alloy 600 (highly Ni enriched metal) were only observed at 337 C (640 F) and only along the boundaries of deformation induced fine grains and subcells. Specimens having deformation free surfaces exhibited continuous uniform inner chromite spinel oxide layers. Specimens with machining induced surface deformation produced non-uniform inner layer oxides (chromite spinel, Cr{sub 2}O{sub 3} and unoxidized material). PWSCC crack tip oxides, in contrast, were fine grain (no duplex structure) and consisted of both chromium rich spinels and ''NiO'' structure oxides. Generally, nickel rich oxides were more abundant under more oxidized conditions (reduced coolant hydrogen) and spinel rich crack tip oxides were favored under more reducing conditions (increased coolant hydrogen). Bulk surface corrosion film thickness did not correlate with observed SCC growth rates. These results suggest that corrosion is not the rate controlling step of PWSCC but rather that PWSCC and corrosion have a common rate

  19. Phytoplankton absorption predicts patterns in primary productivity in Australian coastal shelf waters

    Science.gov (United States)

    Robinson, C. M.; Cherukuru, N.; Hardman-Mountford, N. J.; Everett, J. D.; McLaughlin, M. J.; Davies, K. P.; Van Dongen-Vogels, V.; Ralph, P. J.; Doblin, M. A.

    2017-06-01

    The phytoplankton absorption coefficient (aPHY) has been suggested as a suitable alternate first order predictor of net primary productivity (NPP). We compiled a dataset of surface bio-optical properties and phytoplankton NPP measurements in coastal waters around Australia to examine the utility of an in-situ absorption model to estimate NPP. The magnitude of surface NPP (0.20-19.3 mmol C m-3 d-1) across sites was largely driven by phytoplankton biomass, with higher rates being attributed to the microplankton (>20 μm) size class. The phytoplankton absorption coefficient aPHY for PAR (photosynthetically active radiation; āPHY)) ranged from 0.003 to 0.073 m-1, influenced by changes in phytoplankton community composition, physiology and environmental conditions. The aPHY coefficient also reflected changes in NPP and the absorption model-derived NPP could explain 73% of the variability in measured surface NPP (n = 41; RMSE = 2.49). The absorption model was applied to two contrasting coastal locations to examine NPP dynamics: a high chlorophyll-high variation (HCHV; Port Hacking National Reference Station) and moderate chlorophyll-low variation (MCLV; Yongala National Reference Station) location in eastern Australia using the GIOP-DC satellite aPHY product. Mean daily NPP rates between 2003 and 2015 were higher at the HCHV site (1.71 ± 0.03 mmol C m-3 d-1) with the annual maximum NPP occurring during the austral winter. In contrast, the MCLV site annual NPP peak occurred during the austral wet season and had lower mean daily NPP (1.43 ± 0.03 mmol C m-3 d-1) across the time-series. An absorption-based model to estimate NPP is a promising approach for exploring the spatio-temporal dynamics in phytoplankton NPP around the Australian continental shelf.

  20. 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

  1. Assessment of water consumptions in small mediterranean islands' primary schools by means of a long-term online monitoring

    Science.gov (United States)

    Ferraris, Marco; De Gisi, Sabino; Farina, Roberto

    2016-09-01

    A key challenge of our society is improving schools through the sustainable use of resources especially in countries at risk of desertification. The estimation of water consumption is the starting point for the correct dimensioning of water recovery systems. To date, unlike the energy sector, there is a lack of scientific information regarding water consumption in school buildings. Available data refer roughly to indirect estimates by means of utility bills and therefore no information on the role of water leakage in the internal network of the school is provided. In this context, the aim of the work was to define and implement an on-line monitoring system for the assessment of water consumptions in a small Mediterranean island primary school to achieve the following sub-goals: (1) definition of water consumption profile considering teaching activities and secretarial work; (2) direct assessment of water consumptions and leakages and, (3) quantification of the behaviour parameters. The installed monitoring system consisted of 33 water metres (3.24 persons per water metre) equipped with sensors set on 1-L impulse signal and connected to a data logging system. Results showed consumptions in the range 13.6-14.2 L/student/day and leakage equal to 54.8 % of the total water consumptions. Considering the behavioural parameters, the consumptions related to toilet flushing, personal, and building cleaning were, respectively, 54, 43 and 3 % of the total water ones. Finally, the obtained results could be used for dimensioning the most suitable water recovery strategies at school level such as grey water or rainwater recovery systems.

  2. Effects of sewage water on bio-optical properties and primary production of coastal systems in West Australia

    DEFF Research Database (Denmark)

    Stæhr, Peter Anton; Waite, A. M.; Markager, S.

    2008-01-01

    pigments, 20-50% higher Chl a-specific absorption coefficients and higher photosynthetic capacity. Although maximum rates of Chl a-normalised photosynthesis were strongly related to nitrate availability, no effects were found on the derived areal primary production or algal biomass suggesting......Relationships between key phytoplankton attributes including Chl a-specific light absorption, pigment composition and concentration, photosynthesis, primary production and community structure were studied in two open shallow nutrient-poor coastal systems receiving similar amounts of sewage water...... that photosynthetic and optical parameters are more sensitive indicators of nutrient enrichment than biomass or productivity....

  3. 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)

  4. 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)

  5. 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.

  6. 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)

  7. 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.

  8. 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.

  9. An exploration of factors that influence the regular consumption of water by Irish primary school children.

    LENUS (Irish Health Repository)

    Molloy, C Johnston

    2008-10-01

    Inadequate hydration has been linked to many factors that may impact on children\\'s education and health. Teachers play an important role in the education and behaviour of children. Previous research has demonstrated low water intake amongst children and negative teachers\\' attitudes to water in the classroom. The present study aimed to explore teachers\\' knowledge about water and the perceived barriers to allowing children access to water during lesson time.

  10. Evaluation of chemical phenomena that could have an effect on the performance of recirculation strainers in a Ringhals PWR; Bedoemning av kemiska fenomen med betydelse foer silfunktionen i Ringhals PWR-reaktorer

    Energy Technology Data Exchange (ETDEWEB)

    Liljenzin, Jan-Olov [Liljenzins data och kemikonsult, Goeteborg (Sweden)

    2005-01-15

    An evaluation has been made of the various chemical phenomena that could have an effect on the performance of recirculation strainers after a LOCA in a PWR. Values of pH and concentrations in the water at the bottom of the containment have been calculated as functions of time and temperature for a postulated LOCA. The behaviour of glass wool insulation, its dissolution, and precipitation of amorphous silic acid have been evaluated. Also the corrosion of galvanized surfaces has been considered. Dissolution of zinc by hot boric acid solution can lead to a later precipitation of amorphous zinc hydroxide or phosphate when pH increases and temperature drops. Also a possible growth of microorganisms is discussed. A rough classification of the various phenomena possible along a simplified time scale yields the following conclusions: Hours after the beginning of the LOCA: Precipitation of zinc hydroxide and/or phosphate. Dissolution of glass wool giving rise to an increasing concentration of silic acid in the water. Days after the beginning of the LOCA: Continued dissolution of glass wool and increasing concentration of silica in the water. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Weeks after the beginning of the LOCA: Continued slow dissolution of glass wool leading to a risk of precipitation of amorphous silica. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Initial growth of microorganisms in the water and on surfaces after mutations and adaptation to the existing environment. Months after the beginning of the LOCA: Continued slow dissolution of glass wool leading to a risk of precipitation of amorphous silica. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Continued growth of adapted microorganisms.

  11. 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

  12. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  13. 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.

  14. Preliminary study for extension and improvement on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aly, Omar F.; Mattar Neto, Miguel M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: ofaly@ipen.br, e-mail: mmattar@ipen.br; Schvartzman, Monica M.M.A.M. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: monicas@cdtn.br

    2009-07-01

    This study is for to extend, to improve the existing models, and to propose a local approach to assess the primary water stress corrosion cracking in nickel-based components. It is includes a modeling of new data for Alloy 182 and new considerations about initiation and crack growth according a developing method based on EPRI-MRP-115 (2004), and USNRC NUREG/CR-6964 (2008). The experimental data is obtained from CDTN-Brazilian Nuclear Technology Development Center, by tests through slow strain rate test (SSRT) equipment. The model conception assumed is a built diagram which indicates a thermodynamic condition for the occurrence of corrosion submodes in essayed materials, through Pourbaix diagrams, for Nickel Alloys in high temperature primary water. Over them, are superimposed different models, including a semi-empiric-probabilistic one to quantify the primary water stress corrosion cracking susceptibility, and a crack growth model. These constructed models shall be validated with the experimental data. This development aims to extent some of the models obtained to weld metals like the Alloy 182, and to improve the originals obtained according methodologies exposed in above referred reports. These methodologies comprise laboratory testing procedures, data collecting, data screening, modeling procedures, assembling of data from some laboratories in the world, plotting of results, compared analysis and discussion of these results. Preliminary results for Alloy 182 will be presented. (author)

  15. Assessment of primary production and optical variability in shelf and slope waters near Cape Hatteras, North Carolina. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Redalje, Donald G.; Lohrenz, Stevern E.

    2001-02-12

    In this project we determined primary production and optical variability in the shelf and slope waters off of Cape Hatteras, N.C. These processes were addressed in conjunction with other Ocean Margins Program investigators, during the Spring Transition period and during Summer. We found that there were significant differences in measured parameters between Spring and Summer, enabling us to develop seasonally specific carbon production and ecosystem models as well as seasonal and regional algorithm improvements for use in remote sensing applications.

  16. Artificial primary marine aerosol production: a laboratory study with varying water temperature, salinity, and succinic acid concentration

    Directory of Open Access Journals (Sweden)

    J. Zábori

    2012-11-01

    Full Text Available Primary marine aerosols are an important component of the climate system, especially in the remote marine environment. With diminishing sea-ice cover, better understanding of the role of sea spray aerosol on climate in the polar regions is required. As for Arctic Ocean water, laboratory experiments with NaCl water confirm that a few degrees change in the water temperature (Tw gives a large change in the number of primary particles. Small particles with a dry diameter between 0.01 μm and 0.25 μm dominate the aerosol number density, but their relative dominance decreases with increasing water temperature from 0 °C where they represent 85–90% of the total aerosol number to 10 °C, where they represent 60–70% of the total aerosol number. This effect is most likely related to a change in physical properties and not to modification of sea water chemistry. A change of salinity between 15 g kg−1 and 35 g kg−1 did not influence the shape of a particle number size distribution. Although the magnitude of the size distribution for a water temperature change between 0 °C and 16 °C changed, the shape did not. An experiment where succinic acid was added to a NaCl water solution showed, that the number concentration of particles with 0.010 μm < Dp < 4.5 μm decreased on average by 10% when the succinic acid concentration in NaCl water at a water temperature of 0 °C was increased from 0 μmol L−1 to 94 μmol L−1. A shift to larger sizes in the particle number size distribution is observed from pure NaCl water to Arctic Ocean water. This is likely a consequence of organics and different inorganic salts present in Arctic Ocean water in addition to the NaCl.

  17. Artificial primary marine aerosol production: a laboratory study with varying water temperature, salinity and succinic acid concentration

    Directory of Open Access Journals (Sweden)

    J. Zábori

    2012-08-01

    Full Text Available Primary marine aerosols are an important component of the climate system, especially in the remote marine environment. With diminishing sea-ice cover, better understanding of the role of sea spray aerosol on climate in the polar regions is required. As for Arctic Ocean water, laboratory experiments with NaCl water confirm that a few degrees change in the water temperature (Tw gives a large change in the number of primary particles. Smaller particles with a dry diameter between 0.01 μm and 0.25 μm dominate the aerosol number density, but their relative dominance decreases with increasing water temperature from 0 °C where they represent 85–90% of the total aerosol number to 60–70% of the total aerosol number at 10 °C water temperature. This effect is most likely related to a change in physical properties and not to modification of sea water chemistry. A change of salinity between 15 g kg−1 and 35 g kg−1 showed no influence on the relative shape of a particle number size distribution, nor did a change in water temperature between 0 °C and 16 °C. An experiment where succinic acid was added to a NaCl water solution showed, that the number concentration of particles with Dp < 0.312 μm decreased by 43% when the succinic acid concentration in NaCl water at Tw = 0 °C was increased from 0 μmol l−1 to 2446 μmol l−1. Different organic constituents and perhaps inorganic substances resulted in a particle number shift towards larger particle sizes, when comparing a size distribution resulting from pure NaCl water to size distributions resulting from Arctic Ocean water and resulting from NaCl water with a succinic acid concentration of 2446 μmol l−1.

  18. 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.

  19. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Clerc, T.; Hebert, A. [Institut de Genie Nucleaire, Station Centre-Ville, Montreal, QC, H3C 3A7 (Canada); Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B. [Electricite de France, R and D, SINETICS, 1 Av. du General de Gaulle, 92141, Clamart (France)

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  20. Large Scale Finite Element Thermal Analysis of the Bolts of a French PWR Core Internal Baffle Structure

    Energy Technology Data Exchange (ETDEWEB)

    Rupp, Isabelle; Christophe, Peniguel [EDF R and D, Paris (France); Tommy, Martin Michel [1 av du General de Gaulle, Paris (France)

    2009-11-15

    The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The Electricite De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code{sub S}aturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer

  1. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  2. A Study of Neutronics Effects of the Spacer Grids in a Typical PWR via Monte Carlo Calculation

    Directory of Open Access Journals (Sweden)

    Xuan Bach Tran

    2016-02-01

    Full Text Available Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR. The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400 core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the “volume-preserving” streamlined heterogeneous spacer grids, but the “banded” dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic analysis.

  3. 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.

  4. Free Available Chlorine Disinfection Criteria for Fixed Army Installation Primary Drinking Water

    Science.gov (United States)

    1981-12-01

    disease Disinfection criteria Potable water Enterobacteria Water treatment 20. 3&TACT(een~usm r v fv . it ,ecossoy asd Identify, by b:ock num~ber...in floccu•ation durinF, water treatment.? Oliver aad Visserl0 have shown that fulvic acld fractions of aquatic humic material are n•jor chloroform...already infected bv a virulent organism, or lacks proper immune response, as in cancer therapy . Other pathogenic yeasts and fungi are transmitted to man

  5. 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,

  6. Using polyatomic primary ions to probe an amino acid and a nucleic base in water ice

    Energy Technology Data Exchange (ETDEWEB)

    Conlan, X.A. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: x.conlan@postgrad.manchester.ac.uk; Biddulph, G.X. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: G.Biddulph@postgrad.manchester.ac.uk; Lockyer, N.P. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom); Vickerman, J.C. [Surface Analysis Research Centre, School of Chemical Engineering and Analytical Science, University of Manchester, P.O. Box 88, Manchester M60 1QD (United Kingdom)]. E-mail: John.Vickerman@manchester.ac.uk

    2006-07-30

    In this study on pure water ice, we show that protonated water species [H{sub 2}O] {sub n}H{sup +} are more prevalent than (H{sub 2}O) {sub n} {sup +} ions after bombardment by Au{sup +} monoatomic and Au{sub 3} {sup +} and C{sub 60} {sup +} polyatomic projectiles. This data also reveals significant differences in water cluster yields under bombardment by these three projectiles. The amino acid alanine and the nucleic base adenine in solution have been studied and have been shown to have an effect on the water cluster ion yields observed using an Au{sub 3} {sup +} ion beam.

  7. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Bolt, S.E.

    1977-11-04

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions.

  8. Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-04-01

    In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

  9. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  10. Water footprint of bio-energy and other primary energy carriers

    NARCIS (Netherlands)

    Gerbens-Leenes, P.W.; Hoekstra, A.Y.; Meer, van der Th.H.

    2008-01-01

    Freshwater is essential for life on earth, not only for basic human needs such as food, fibre and drinking water, but also for a healthy environment. In the near future, important challenges are to meet basic needs and to ensure that the extraction of water does not affect freshwater ecosystems. At

  11. WILLING HORSES: THERE IS NO WATER PRIMARY TEACHER EDUCATION IN NEPAL

    Directory of Open Access Journals (Sweden)

    Dr. T. Karunakaran

    2013-07-01

    Full Text Available This article is an outcome of a mini research conducted among 15 primary teachers who are currently teaching English at the primary level in the state funded schools in Kathmandu in Nepal. Basically, it deals with the deplorable condition of the primary teacher education in the country. The main focus of the study is on the English language teachers’ thirst for their professional as well as English language development which has indeed confined to a bottle-neck situation for many reasons. The finding shows that primary English teachers are much willing to develop their language skills but their interests are frozen to death since there is no one to listen to their demands. It has been found that the pre service trainings has been of no use since the subsequent placement is a nightmare and the in service training never focuses on language development domain. Further, it was found that neither the trainers nor the supervisors assist language teachers after the training. On the whole, the training seems to be merely a formality rather than it being intended for the development of the primary level English language teachers.

  12. Evidence for dissolved organic matter as the primary source and sink of photochemically produced hydroxyl radical in arctic surface waters.

    Science.gov (United States)

    Page, Sarah E; Logan, J Robert; Cory, Rose M; McNeill, Kristopher

    2014-04-01

    Hydroxyl radical (˙OH) is an indiscriminate oxidant that reacts at near-diffusion-controlled rates with organic carbon. Thus, while ˙OH is expected to be an important oxidant of dissolved organic matter (DOM) and other recalcitrant compounds, the role of ˙OH in the oxidation of these compounds in aquatic ecosystems is not well known due to the poorly constrained sources and sinks of ˙OH, especially in pristine (unpolluted) natural waters. We measured the rates of ˙OH formation and quenching across a range of surface waters in the Arctic varying in concentrations of expected sources and sinks of ˙OH. Photochemical formation of ˙OH was observed in all waters tested, with rates of formation ranging from 2.6 ± 0.6 to 900 ± 100 × 10(-12) M s(-1). Steady-state concentrations ranged from 2 ± 1 to 290 ± 60 × 10(-17) M, and overlapped with previously reported values in surface waters. While iron-mediated photo-Fenton reactions likely contributed to the observed ˙OH production, several lines of evidence suggest that DOM was the primary source and sink of photochemically produced ˙OH in pristine arctic surface waters. DOM from first-order or headwater streams was more efficient in producing ˙OH than what has previously been reported for DOM, and ˙OH formation decreased with increasing residence time of DOM in sunlit surface waters. Despite the ubiquitous formation of ˙OH in arctic surface waters observed in this study, photochemical ˙OH formation was estimated to contribute ≤4% to the observed photo-oxidation of DOM; however, key uncertainties in this estimate must be addressed before ruling out the role of ˙OH in the oxidation of DOM in these waters.

  13. Primary description of surface water phytoplankton pigment patterns in the Bay of Bengal

    Digital Repository Service at National Institute of Oceanography (India)

    Naik, R.K.; Anil, A.C.; Narale, D.D.; Chitari, R.R.; Kulkarni, V.V.

    . Biogeochemical controls and feedbacks on ocean primary production. Science 281, 200-206. Gibb, S.W., Barlow, R.G., Cummings, D.G., Rees, N.W., Trees, C.C., Holligan, P., Suggett, D., 2000. Surface phytoplankton pigment distributions in the Atlantic Ocean...

  14. Tiniest primary producers in the marine environment: An appraisal from the context of waters around India

    Digital Repository Service at National Institute of Oceanography (India)

    Mitbavkar, S.; Anil, A.C.

    Phytoplankton (0.2 mm - 2 mm) are the major primary producers in the marine environment thereby forming a basic link in the marine food web. They are categorized into different groups depending on their size range. Cells in the size range of 0...

  15. Concerted changes in N and C primary metabolism in alfalfa (Medicago sativa) under water restriction

    OpenAIRE

    Aranjuelo, Iker; Tcherkez, Guillaume,; Molero, Gemma; Gilard, Françoise; AVICE, JEAN-CHRISTOPHE; Nogués, Salvador

    2013-01-01

    Although the mechanisms of nodule N2 fixation in legumes are now well documented, some uncertainty remains on the metabolic consequences of water deficit. In most cases, little consideration is given to other organs and, therefore, the coordinated changes in metabolism in leaves, roots, and nodules are not well known. Here, the effect of water restriction on exclusively N2-fixing alfalfa (Medicago sativa L.) plants was investigated, and proteomic, metabolomic, and physiological analyses were ...

  16. Imaging of primary and secondary radiation-Modelling and experimental results of a radioactive source and a water phantom

    Science.gov (United States)

    Gamage, K. A. A.; Taylor, G. C.; Joyce, M. J.

    2014-11-01

    In this paper the contribution of primary and secondary radiation from a water phantom to a pinhole volume, as a result of three neutron sources (Cf, AmBe and 5 MeV mono-energetic) and two gamma sources (Cs and Co), is separately estimated using the PTRAC particle tracking option available in MCNP. Also in this paper imaging of the mixed radiation field produced by a Van de Graaf accelerator (when a water phantom is present) is described. In the model, a spherical tally volume, 2 cm in diameter, was placed equidistantly from a radioactive source and 30×30×15 cm3 water phantom. Monte Carlo simulations have been carried out to investigate the level of primary and secondary radiation contributing to the pinhole volume directly from the source and from interactions in the phantom respectively. The spatial distribution of counts clearly discriminated the source and the phantom. The results have shown that the percentage of neutrons reflected from the phantom with energies above 1 MeV increases with mean energy of the source. This method has significant potential to characterise secondary radiation in proton therapy, where it would help to verify the location and the energy delivered during the treatment.

  17. 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.

  18. Prevalence of dental fluorosis among primary school children in association with different water fluoride levels in Mysore district, Karnataka.

    Science.gov (United States)

    Sebastian, Shibu Thomas; Soman, Rino Roopak; Sunitha, S

    2016-01-01

    Fluoride intake at optimal level decreases the incidence of dental caries. However, excessive intake, especially during developmental stages can cause adverse effects such as dental and skeletal fluorosis. To assess the prevalence and severity of dental fluorosis in primary school children born and raised in three villages of Mysore District. The three selected villages have different water fluoride concentrations. Three villages namely, Nerale (water fluoride 2.0 ppm), Belavadi (1.2 ppm) and Naganahally (0.4 ppm) were selected for the study. Then, a total of 405 children, 10-12-year-old (204 [50.4%] males and 201 [49.60%] females) were selected from three schools of the villages. Dean's fluorosis index recommended by World Health Organization was used to evaluate fluorosis among the study population. The overall prevalence of dental fluorosis was found to be 41.73%. An increase in the community fluorosis index (CFI) was higher among those living in high water fluoride area. A significantly positive correlation was found between CFI and water fluoride concentration in drinking water.

  19. Concerted changes in N and C primary metabolism in alfalfa (Medicago sativa) under water restriction.

    Science.gov (United States)

    Aranjuelo, Iker; Tcherkez, Guillaume; Molero, Gemma; Gilard, Françoise; Avice, Jean-Christophe; Nogués, Salvador

    2013-02-01

    Although the mechanisms of nodule N(2) fixation in legumes are now well documented, some uncertainty remains on the metabolic consequences of water deficit. In most cases, little consideration is given to other organs and, therefore, the coordinated changes in metabolism in leaves, roots, and nodules are not well known. Here, the effect of water restriction on exclusively N(2)-fixing alfalfa (Medicago sativa L.) plants was investigated, and proteomic, metabolomic, and physiological analyses were carried out. It is shown that the inhibition of nitrogenase activity caused by water restriction was accompanied by concerted alterations in metabolic pathways in nodules, leaves, and roots. The data suggest that nodule metabolism and metabolic exchange between plant organs nearly reached homeostasis in asparagine synthesis and partitioning, as well as the N demand from leaves. Typically, there was (i) a stimulation of the anaplerotic pathway to sustain the provision of C skeletons for amino acid (e.g. glutamate and proline) synthesis; (ii) re-allocation of glycolytic products to alanine and serine/glycine; and (iii) subtle changes in redox metabolites suggesting the implication of a slight oxidative stress. Furthermore, water restriction caused little change in both photosynthetic efficiency and respiratory cost of N(2) fixation by nodules. In other words, the results suggest that under water stress, nodule metabolism follows a compromise between physiological imperatives (N demand, oxidative stress) and the lower input to sustain catabolism.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  5. 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)

  6. Relationships between primary productivity and bottom-water oxygenation off northwest Africa during the last deglaciation

    NARCIS (Netherlands)

    Filipsson, H.L.; Romero, O.E.; Stuut, J.B.W.

    2011-01-01

    The upwelling region off northwest Africa is one of the most productive regions in the world ocean. This study details the response of surface-and deep-water environments off Mauritania, northwest Africa, to the rapid climate events of the last deglaciation, especially the Bolling-Allerod (15.5-13.5

  7. 75 FR 15499 - National Primary Drinking Water Regulations; Announcement of the Results of EPA's Review of...

    Science.gov (United States)

    2010-03-29

    ... 20460. Hand Delivery: EPA Docket Center Public Reading Room, EPA Headquarters West, Room 3334, 1301...--hexavalent chromium CWS--community water system DBPs--disinfection byproducts DBCP--1,2-dibromo-3-chloropropane DBPR--Disinfectants and Disinfection Byproducts Rule DEHA--di(2-ethylhexyl)adipate...

  8. The Design and Development of a Simulation to Teach Water Conservation to Primary School Students

    Science.gov (United States)

    Campbell, Lee

    2004-01-01

    Information and Communications Technology (ICT) plays a dominant role in enhancing teaching and learning. Similar advances have been made in the use of multimedia in the classroom. These advances are coupled with newer developmental tools and techniques. This paper examines the design and development of a simulation on water conservation. Science…

  9. 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.)

  10. 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).

  11. 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.

  12. 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.

  13. 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.

  14. Gravitropic reaction of primary seminal roots of Zea mays L. influenced by temperature and soil water potential.

    Science.gov (United States)

    Nakamoto, T

    1995-03-01

    The growth of the primary seminal root of maize (Zea mays L.) is characterized by an initial negative gravitropic reaction and a later positive one that attains a plagiotropic liminal angle. The effects of temperature and water potential of the surrounding soil on these gravitropic reactions were studied. Temperatures of 32, 25, and 18C and soil water potentials of -5, -38, and -67 kPa were imposed and the direction of growth was measured for every 1 cm length of the root. The initial negative gravitropic reaction extended to a distance of about 10 cm from the grain. Higher temperatures reduced the initial negative gravitropic reaction. Lower soil water potential induced a downward growth at root emergence. A mathematical model, in which it was assumed that the rate of the directional change of root growth was a sum of a time-dependent negative gravitropic reaction and an establishment of the liminal angle, adequately fitted the distance-angle relations. It was suggested that higher temperatures and/or a lower water potential accelerated the diminution of the initial negative gravitropic reaction.

  15. 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.

  16. Axial gas flow in irradiated PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data.

  17. Water Clear Cell Adenoma of the Parathyroid Gland: A Forgotten Cause of Primary Hyperparathyroidism.

    Science.gov (United States)

    El Hussein, Siba; Poppiti, Robert

    2017-08-01

    Water clear cell adenoma of the parathyroid gland is a rare neoplasm composed of large clear cells with foamy cytoplasm and mild nuclear pleomorphism, compressing the residual nonneoplastic parathyroid tissue. The differential diagnosis includes a variety of neoplasms with clear cell features. In this article, we provide an overview of the entity with a historical perspective, in order to help pathologists in distinguishing it from other neoplasms in their daily practice.

  18. Primary tuberculosis of urethra presenting as stricture urethra and watering can perineum: A rarity.

    Science.gov (United States)

    Prakash, Gaurav; Singh, Vishwajeet; Sinha, Rahul Janak; Babu, Suresh; Jhanwar, Ankur; Mehrotra, C N

    2016-01-01

    A young man presented with irritative lower urinary tract symptoms and multiple fistulae (watering can) in the perineum since 6 months. Micturating cystourethrogram and retrograde urethrogram was performed after 12 weeks following suprapubic cystostomy which showed bulbar urethral stricture with multiple urethrocutaneous fistulae. He underwent anastomotic urethroplasty and excision of the urethrocutaneous fistulae. Histopathology of the excised fistulous tract showed granulomatous pathology suggestive of tuberculosis. Antitubercular treatment was given for 9 months. The patient is voiding well at 12 months follow-up.

  19. Primary factors affecting water quality and quantity in four watersheds in Eastern Puerto Rico

    Science.gov (United States)

    Murphy, Sheila F.; Stallard, Robert F.

    2009-01-01

    As part of the U.S. Geological Survey (USGS) Water, Energy, and Biogeochemical Budgets (WEBB) program, four small watersheds in eastern Puerto Rico were monitored to identify and evaluate the effects of geology, landcover, atmospheric deposition, and other factors on stream water quality and quantity. Two catchments are located on coarse-grained granitic plutonic rocks, which weather to quartz- and clay-rich, sandy soils, and two are located on fine-grained volcanic rocks and volcaniclastic sediments, which weather to quartz-poor, fine-grained soils. These differing soil materials result in different hydrologic regimes. Soils on the granitic rocks have greater permeability than those developed on the volcaniclastic rocks, allowing more water infiltration and potentially greater landslide erosion rates. For each bedrock type, one catchment was covered with mature rainforest, and the other catchment was affected by agricultural practices typical of eastern Puerto Rico. These practices led to the erosion of much of the original surface soil in the agricultural watersheds, which introduced large quantities of sediment to stream channels. The agricultural watersheds are undergoing natural reforestation, like much of Puerto Rico. Eastern Puerto Rico receives large atmospheric inputs of marine salts, pollutants from the Northern Hemisphere, and Saharan Desert dust. Marine salts contribute over 80 percent of the ionic charge in precipitation, with peak inputs in January. Intense storms, mostly hurricanes, are associated with exceptionally high chloride concentrations in stream waters. Temperate pollution contributes nitrate, ammonia, and sulfate, with maximum inputs during northern cold fronts in January, April, and May. Pollution inputs have increased through time. Desert dust peaks in June and July, during times of maximum dust transport from the Saharan Desert across the Atlantic Ocean.

  20. Hydrogen-enriched water restoration of impaired calcium propagation by arsenic in primary keratinocytes

    Science.gov (United States)

    Yu, Wei-Tai; Chiu, Yi-Ching; Lee, Chih-Hung; Yoshioka, Tohru; Yu, Hsin-Su

    2013-11-01

    Endemic contamination of artesian water for drinking by arsenic is known to cause several human cancers, including cancers of the skin, bladder, and lungs. In skin, multiple arsenic-induced Bowen's disease (As-BD) can develop into invasive cancers after decades of arsenic exposure. The characteristic histological features of As-BD include full-layer epidermal dysplasia, apoptosis, and abnormal proliferation. Calcium propagation is an essential cellular event contributing to keratinocyte differentiation, proliferation, and apoptosis, all of which occur in As-BD. This study investigated how arsenic interferes calcium propagation of skin keratinocytes through ROS production and whether hydrogen-enriched water would restore arsenic-impaired calcium propagation. Arsenic was found to induce oxidative stress and inhibit ATP- and thapsigaragin-induced calcium propagation. Pretreatment of arsenic-treated keratinocytes by hydrogen-enriched water or beta-mercaptoethanol with potent anti-oxidative effects partially restored the propagation of calcium by ATP and by thapsigaragin. It was concluded that arsenic may impair calcium propagation, likely through oxidative stress and interactions with thiol groups in membrane proteins.

  1. Direct energy recovery from primary and secondary sludges by supercritical water oxidation.

    Science.gov (United States)

    Svanström, M; Modell, M; Tester, J

    2004-01-01

    Supercritical water oxidation (SCWO) oxidizes organic and biological materials virtually completely to benign products without the need for stack gas scrubbing. Heavy metals are recovered as stabilized solid, along with the sand and clay that is present in the feed. The technology has been under development for twenty years. The major obstacle to commercialization has been developing reactors that are not clogged by inorganic solid deposits. That problem has been solved by using tubular reactors with fluid velocities that are high enough to keep solids in suspension. Recently, system designs have been created that reduce the cost of processing sewage sludges below that of incineration. At 10 wt- % dry solids, sludge can be oxidized with virtually complete recovery of the sludge heating value as hot water or high-pressure steam. Liquid carbon dioxide of high purity can be recovered from the gaseous effluent and excess oxygen can be recovered for recycle. The net effect is to reduce the stack to a harmless vent with minimal flow rate of a clean gas. Complete simulations have been developed using physical property models that accurately simulate the thermodynamic properties of sub- and supercritical water in mixtures with O2, N2, CO2, and organics. Capital and operating cost estimates are given for sewage sludge treatment, which are less costly than incineration. The scenario of direct recovery of energy from sludges has inherent benefits compared to other gasification or liquefaction options.

  2. Direct Energy Recovery from Primary and Secondary Sludges by Supercritical Water Oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Svanstroem, Magdalena; Modell, Michael; Tester, Jefferson

    2003-07-01

    Supercritical water oxidation (SCWO) oxidizes organic and biological materials virtually completely to benign products without the need for stack gas scrubbing. Heavy metals are recovered as stabilized solid, along with the sand and clay that is present in the feed. The technology has been under development for twenty years. The major obstacle to commercialization has been developing reactors that are not clogged by inorganic solid deposits. That problem has been solved by using tubular reactors with fluid velocities that are high enough to keep solids in suspension. Recently, system designs have been created that reduce the cost of processing sewage sludges below that of incineration. At 10 wt-% dry solids, sludge can be oxidized with virtually complete recovery of the sludge heating value as hot water or high-pressure steam. Liquid carbon dioxide of high purity can be recovered from the gaseous effluent and excess oxygen can be recovered for recycle. The net effect is to reduce the stack to a harmless vent with minimal flow rate of a clean gas. Complete simulations have been developed using physical property models that accurately simulate the thermodynamic properties of sub- and supercritical water in mixtures with O{sub 2}, N{sub 2}, CO{sub 2}. and organics. Capital and operating cost estimates are given for sewage sludge treatment, which are less costly than incineration. The scenario of direct recovery of energy from sludges has inherent benefits compared to other gasification or liquefaction options. (author)

  3. Determination of ammonia and primary amine compounds and Kjeldahl nitrogen in water samples with a modified Roth's fluorimetric method.

    Science.gov (United States)

    Lloret, S Meseguer; Andrés, J Verdú; Legua, C Molins; Falcó, P Campíns

    2005-02-28

    A method for the simultaneous determination of primary amino groups and ammonium ion has been proposed. The method is based in solution derivatization with o-Phthaldialdehyde/N-acetyl-cisteine (OPA/NAC) and fluorescence measurement of the formed isoindols. Analytical characteristics and description of the developed procedure have been provided. The calibration graphs for ammonium (up to 1.44mgL(-1) of N) and methylamine as primary amino model compound (up to 0.282mgL(-1) of N), were obtained. Bivariate and multivariate calibration models have been tested. The limits of detection were 0.07mgL(-1) of N and 0.004mgL(-1) of N for ammonium and amine, respectively. The procedure was first applied directly to standard solutions containing ammonium and amines and secondly to digested solutions by Kjeldahl method. The results obtained allowed to establish the best digestion conditions in order to perform the total amine conversion into ammonium. This procedure has been also applied to real samples (irrigation ditch, residual and fountain waters) and the concentrations of primary amine groups and ammonium have been evaluated. The results obtained after Kjeldahl digestion allowed to estimate the total Kjeldahl N contained in the samples. The samples were also analysed by Nessler method and similar results were obtained.

  4. 核电站汽水管道流动加速腐蚀的影响因素分析及对策%Cause Analysis on FAC Failures of Steam/Water Piping in Secondary Loop of PWR Power Plants and the Countermeasures

    Institute of Scientific and Technical Information of China (English)

    张桂英; 顾宇; 邵杰

    2012-01-01

    Targeting at the pipe rupture accidents occurring all over the world in the pressurized water reactor(PWR) secondary loop induced by flow accelerated corrosion(FAC) since 1980s,which have led to economic losses and social impact,experimental studies are carried out to analyze the FAC corrosion behavior,formation mechanism and influencing factors,etc.after which preventive measures and strategies are proposed to deal with the FAC resulted from manual mishandling.This may serve as a reference for design and operation of pressurized water reactor(PWR) power plant in the prevention of FAC failures.%针对20世纪80年代以来国际上压水堆核电站二回路发生的因流动加速腐蚀引起的管道破裂事故造成的经济损失和社会影响,介绍了因流动加速腐蚀引起的事故及其造成的影响,进行了有关流动加速腐蚀特征、形成机理和影响因素方面的研究,并针对可以人为控制的因素提出了预防措施和对策,为压水堆核电站设计和运行人员提供因流动加速腐蚀对核安全造成的危害加以预防提供参考.

  5. An evaluation of design characteristics for the water chemistry for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B. S.; Kim, J. P.; Kim, H. Y.; Yoon, J. H.; Lee, D. J

    1998-07-01

    SMART does not have boron as a chemical shim using boron and does not operate purification system in power operation. However, the water chemistry applied in SMART still needs to meet the integrity and reliable operation requirements. This report reviews the corrosion resistance of the construction materials for SMART and the chemistry operating parameters qualitatively by comparing with those of PWR. The review shows that the construction materials for SMART, such as titanium alloys for steam generator tube materials and zirconium alloy for fuel cladding have good guarantee the integrity and reliability of the primary circuit and fuel cladding as well as the secondary circuit. This review results can be utilized as input data for SMART chemistry design, such as chemistry design guide, and chemistry manual. (author). 25 refs., 8 tabs., 17 figs

  6. 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.

  7. Primary productivity

    Digital Repository Service at National Institute of Oceanography (India)

    Verlecar, X.N.; Parulekar, A.H.

    Photosynthetic production in the oceans in relation to light, nutrients and mixing processes is discussed. Primary productivity in the estuarine region is reported to be high in comparison to coastal and oceanic waters. Upwelling phenomenon...

  8. Primary tuberculosis of urethra presenting as stricture urethra and watering can perineum: A rarity

    Directory of Open Access Journals (Sweden)

    Gaurav Prakash

    2016-01-01

    Full Text Available A young man presented with irritative lower urinary tract symptoms and multiple fistulae (watering can in the perineum since 6 months. Micturating cystourethrogram and retrograde urethrogram was performed after 12 weeks following suprapubic cystostomy which showed bulbar urethral stricture with multiple urethrocutaneous fistulae. He underwent anastomotic urethroplasty and excision of the urethrocutaneous fistulae. Histopathology of the excised fistulous tract showed granulomatous pathology suggestive of tuberculosis. Antitubercular treatment was given for 9 months. The patient is voiding well at 12 months follow-up.

  9. Simulated nutrient dissolution of Asian aerosols in various atmospheric waters: Potential links to marine primary productivity

    Science.gov (United States)

    Wang, Lingyan; Bi, Yanfeng; Zhang, Guosen; Liu, Sumei; Zhang, Jing; Xu, Zhaomeng; Ren, Jingling; Zhang, Guiling

    2017-09-01

    To probe the bioavailability and environmental mobility of aerosol nutrient elements (N, P, Si) in atmospheric water (rainwater, cloud and fog droplets), ten total suspended particulate (TSP) samples were collected at Fulong Mountain, Qingdao from prevailing air mass trajectory sources during four seasons. Then, a high time-resolution leaching experiment with simulated non-acidic atmospheric water (non-AAW, Milli-Q water, pH 5.5) and subsequently acidic atmospheric water (AAW, hydrochloric acid solution, pH 2) was performed. We found that regardless of the season or source, a monotonous decreasing pattern was observed in the dissolution of N, P and Si compounds in aerosols reacted with non-AAW, and the accumulated dissolved curves of P and Si fit a first-order kinetic model. No additional NO3- + NO2- dissolved out, while a small amount of NH4+ in Asian dust (AD) samples was released in AAW. The similar dissolution behaviour of P and Si from non-AAW to AAW can be explained by the Transition State Theory. The sources of aerosols related to various minerals were the natural reasons that affected the amounts of bioavailable phosphorus and silicon in aerosols (i.e., solubility), which can be explained by the dissolution rate constant of P and Si in non-AAW with lower values in mineral aerosols. The acid/particle ratio and particle/liquid ratio also have a large effect on the solubility of P and Si, which was implied by Pearson correlation analysis. Acid processing of aerosols may have great significance for marine areas with limited P and Si and post-acidification release increases of 1.1-10-fold for phosphorus and 1.2-29-fold for silicon. The decreasing mole ratio of P and Si in AAW indicates the possibility of shifting from a Si-limit to a P-limit in aerosols in the ocean, which promotes the growth of diatoms prior to other algal species.

  10. First international comparison of primary absorbed dose to water standards in the medium-energy X-ray range

    Science.gov (United States)

    Büermann, Ludwig; Guerra, Antonio Stefano; Pimpinella, Maria; Pinto, Massimo; de Pooter, Jacco; de Prez, Leon; Jansen, Bartel; Denoziere, Marc; Rapp, Benjamin

    2016-01-01

    This report presents the results of the first international comparison of primary measurement standards of absorbed dose to water for the medium-energy X-ray range. Three of the participants (VSL, PTB, LNE-LNHB) used their existing water calorimeter based standards and one participant (ENEA) recently developed a new standard based on a water-graphite calorimeter. The participants calibrated three transfer chambers of the same type in terms of absorbed dose to water (NDw) and in addition in terms of air kerma (NK) using the CCRI radiation qualities in the range 100 kV to 250 kV. The additional NK values were intended to be used for a physical analysis of the ratios NDw/NK. All participants had previously participated in the BIPM.RI(I)-K3 key comparison of air kerma standards. Ratios of pairs of NMI's NK results of the current comparison were found to be consistent with the corresponding key comparison results within the expanded uncertainties of 0.6 % - 1 %. The NDw results were analysed in terms of the degrees of equivalence with the comparison reference values which were calculated for each beam quality as the weighted means of all results. The participant's results were consistent with the reference value within the expanded uncertainties. However, these expanded uncertainties varied significantly and ranged between about 1-1.8 % for the water calorimeter based standards and were estimated at 3.7 % for the water-graphite calorimeter. It was shown previously that the ratios NDw/NK for the type of ionization chamber used as transfer chamber in this comparison were very close (within less than 1 %) to the calculated values of (bar muen/ρ)w,ad, the mean values of the water-to-air ratio of the mass-energy-absorption coefficients at the depth d in water. Some of the participant's results deviated significantly from the expected behavior. Main text To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of

  11. 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)

  12. 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。

  13. Impacts of climate and land use change on ecosystem hydrology and net primary productivity: Linking water availability to food security in Asia

    Science.gov (United States)

    Dangal, S. R. S.; Tian, H.; Pan, S.; Zhang, B.; Yang, J.

    2015-12-01

    The nexus approach to food, water and energy security in Asia is extremely important and relevant as the region has to feed two-third of the world's population and accounts for 59% of the global water consumption. The distribution pattern of food, water and energy resources have been shaped by the legacy effect of both natural and anthropogenic disturbances and therefore are vulnerable to climate change and human activities including land use/cover change (LUCC) and land management (irrigation and nitrogen fertilization). In this study, we used the Dynamic Land Ecosystem Model (DLEM) to examine the effects of climate change, land use/cover change, and land management practices (irrigation and nitrogen fertilization) on the spatiotemporal trends and variability in water availability and its role in limiting net primary productivity (NPP) and food security in the 20th and early 21st centuries. Our specific objectives are to quantify how climate change, LUCC and other environmental changes have interactively affected carbon and water dynamics across the Asian region. In particular, we separated the Asian region into several sub-region based on the primary limiting factor - water, food and energy. We then quantified how changes in environmental factors have altered the water and food resources during the past century. We particularly focused on Net Primary Productivity (NPP) and water cycle (Evapotranspiration, discharge, and runoff) as a measure of available food and water resources, respectively while understanding the linkage between food and water resources in Asia.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. PNL technical review of pressurized thermal-shock issues. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J.; Simonen, E.P.; Simonen, F.A.; Stevens, D.L.; Taylor, T.T.

    1982-07-01

    Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.

  1. PWR hot leg natural circulation modeling with MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1997-12-31

    Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)

  2. 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.

  3. Inactivation and injury of total coliform bacteria after primary disinfection of drinking water by TiO2 photocatalysis.

    Science.gov (United States)

    Rizzo, Luigi

    2009-06-15

    In this study the potential application of TiO(2) photocatalysis as primary disinfection system of drinking water was investigated in terms of coliform bacteria inactivation and injury. As model water the effluent of biological denitrification unit for nitrate removal from groundwater, which is characterized by high organic matter and bacteria release, was used. The injury of photocatalysis on coliform bacteria was characterized by means of selective (mEndo) and less selective (mT7) culture media. Different catalyst loadings as well as photolysis and adsorption effects were investigated. Photocatalysis was effective in coliform bacteria inactivation (91-99% after 60 min irradiation time, depending on both catalyst loading and initial density of coliform bacteria detected by mEndo), although no total removal was observed after 60 min irradiation time. The contribution of adsorption mechanism was significant (60-98% after 60 min, depending on catalyst loading) compared to previous investigations probably due to the nature of source water rich in particulate organic matter and biofilm. Photocatalysis process did not result in any irreversible injury (98.8% being the higher injury) under investigated conditions, thus a bacteria regrowth may take place under optimum environment conditions if any final disinfection process (e.g., chlorine or chlorine dioxide) is not used.

  4. Effects of plant diversity on primary production and species interactions in brackish water angiosperm communities

    DEFF Research Database (Denmark)

    Salo, Tiina; Gustafsson, Camilla; Boström, Christoffer

    2009-01-01

    plant productivity in brackish water angiosperm communities, a 14 wk field experiment was conducted. Using a replacement design with a standardized initial aboveground biomass, shoots of Zostera marina, Potamogeton filiformis and P. perfoliatus were planted on a shallow, sandy bottom in replicated...... monocultures and all possible species combinations. Response variables included aboveground and belowground biomass, shoot density, space occupation and porewater nutrients. To determine whether selection and/or complementarity controlled productivity, additive partitioning and Di were calculated. Richness...... effects were species-specific and only increased the biomass production of P. perfoliatus and tuber production of P. filiformis, while species composition generally had a stronger effect on biomass production. Additive partitioning indicated a positive complementarity effect for the aboveground biomass...

  5. 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.

  6. Measurement of water column primary production using photosynthesis-irradiance relations for surface phytoplankton, the vertical chlorophyll profile, and underwater light intensity

    Science.gov (United States)

    Demidov, A. B.; Gagarin, V. I.; Mosharov, S. A.

    2016-09-01

    A method has been developed to measure water column integrated primary production (PPint) in the water column using photosynthesis-irradiance relations for surface phytoplankton, the vertical profile of chlorophyll a concentration, and the underwater light intensity. Good correlation has been found for the results calculated with this method and light dependences in situ. The advantages of this method are the independence of PPint calculation from CTD profiling and water sampling, and thus optimization (reduction) of the station working time.

  7. 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.

  8. Engineering design feasibility of low boron concentration core in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, A. T.; Kim, M. H. [Kyung Hee University, Yongin-shi, Gyeonggi-do, 446-701 Republic of Korea (Korea, Republic of); Woo, I.; Shon, S. R., E-mail: atdaing@khu.ac.k [Korea Nuclear Fuel, 1047 Daedukdaero, Yuseong-gu, Daejeon, 305-353 Republic of Korea (Korea, Republic of)

    2010-10-15

    In pressurized water reactor operation, higher level of soluble boron concentration could contribute higher impact from boron dilution situations, higher amount of liquid waste, and higher radiation dose to operators from higher corrosion potential to cladding and structure. Two practical and feasible means to reduce the maximum boron concentration were investigated in this study. A technically straightforward, possible means, can be achieved either by implementation of enriched boric acid (Eba) or by increasing more shim rod (fixed burnable absorber) worth. A simplest option is that the Eba is applied into reference core (Ref) design, OPR-1000 design, Ulchin unit-5 by allowing use of same fuel assemblies and core design without changing any nuclear design methodology used in that Ref design. Although results of Eba option proved its favorable power distribution and peaking factor, its moderator temperature coefficient (MTC) value reached positive, 3.25 pcm/ C at 40 EFPD which is beyond the design safety limit. An alternative option with more shim rods in fuel assemblies was tried with four types of integral burnable absorbers: gadolinia, integral fuel burnable absorber (Ifba), erbium and alumina boron carbide. Four core design candidates have been developed by keeping major engineering designs and preserving equivalent fuel enrichment level used in Ref design. However, all optimal designs were targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between Ref and optimal core designs is presented here. One of them is suggested as the most promising and favorable low boron core (Lbc) design in this framework. The proper combination of axial and radial enrichment zoning pattern in Lbc design candidate with Ifba-bearing fuel assemblies at equilibrium cycle, could bring 2 times narrower axial offset variation than that of Ref design, and maintain acceptable power peaking factor around 23% lower than

  9. Modelling material effects on flow-accelerated corrosion in primary CANDU coolant and secondary reactor feed-water

    Energy Technology Data Exchange (ETDEWEB)

    Phromwong, P.; Lister, D., E-mail: c7r13@unb.ca [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada); Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2012-07-01

    The effects of chromium content on flow-accelerated corrosion (FAC) of carbon steel have been predicted very well by including a passivating layer, which is a chromium-dependent diffusion barrier at the metal-oxide interface. By adjusting the properties of the chromium-dependent layer, described with a Passivation Parameter (PP), we can predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments (140{sup o}C and neutral or ammoniated chemistry). The model and an appropriate PP are also applied to the environment typical of carbon-steel feeders in the primary coolant of a CANDU reactor (310{sup o}C and lithiated chemistry). The model predicts FAC rate very well (with a deviation of 10% or less) in both situations. (author)

  10. Landscape configuration is the primary driver of impacts on water quality associated with agricultural expansion

    Science.gov (United States)

    Chaplin-Kramer, Rebecca; Hamel, Perrine; Sharp, Richard; Kowal, Virgina; Wolny, Stacie; Sim, Sarah; Mueller, Carina

    2016-07-01

    Corporations and other multinational institutions are increasingly looking to evaluate their innovation and procurement decisions over a range of environmental criteria, including impacts on ecosystem services according to the spatial configuration of activities on the landscape. We have developed a spatially explicit approach and modeled a hypothetical corporate supply chain decision representing contrasting patterns of land-use change in four regions of the globe. This illustrates the effect of introducing spatial considerations in the analysis of ecosystem services, specifically sediment retention. We explored a wide variety of contexts (Iowa, USA; Mato Grosso, Brazil; and Jiangxi and Heilongjiang in China) and these show that per-area representation of impacts based on the physical characterization of a region can be misleading. We found two- to five-fold differences in sediment export for the same amount of habitat conversion within regions characterized by similar physical traits. These differences were mainly determined by the distance between land use changes and streams. The influence of landscape configuration is so dramatic that it can override wide variation in erosion potential driven by physical factors like soil type, slope, and climate. To minimize damage to spatially-dependent ecosystem services like water purification, sustainable sourcing strategies should not assume a direct correlation between impact and area but rather allow for possible nonlinearity in impacts, especially in regions with little remaining habitat and highly variable hydrological connectivity.

  11. Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

    Directory of Open Access Journals (Sweden)

    Li Wang

    2017-02-01

    Full Text Available The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO algorithm that is based on random perturbation (RP-PSO. The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

  12. The effects of nutrient additions on particulate and dissolved primary production in surface waters of three Mediterranean eddies

    Science.gov (United States)

    Lagaria, A.; Psarra, S.; Lefèvre, D.; van Wambeke, F.; Courties, C.; Pujo-Pay, M.; Oriol, L.; Tanaka, T.; Christaki, U.

    2010-12-01

    The effects of additions of nitrogen (+N), phosphorus (+P), alone and in combination, were assessed during three microcosm experiments performed with surface waters of three anticyclonic eddies, located in the Western, Central and Eastern Mediterranean. We examined the effects of nutrient additions on rates of dissolved and particulate primary production and on metabolic rates of the osmotrophic community (phytoplankton and heterotrophic prokaryotes). The experiments were performed in June/July 2008 during the BOUM (Biogeochemistry from the Oligotrophic to the Ultra-oligotrophic Mediterranean) cruise. In all three experiments, particulate primary production was significantly stimulated by the additions of nitrogen (+N, +NP) while no effect was observed with the addition of phosphorus alone. Percent extracellular release (PER) showed an inverse relation with total primary production (PPtotal), displaying the lowest values (4-8%) in the +NP treatment. Among the three treatments, the +NP had the strongest effect on the community metabolic rates leading to positive net community production values (NCP>0). These changes of NCP were mainly due to enhanced gross community production (GCP) rather than lower respiration rates (CR). In +NP treatments autotrophic production (whether expressed as GCP or PPtotal) was high enough to fulfil the carbon requirements of the heterotrophic prokaryotes, with phytoplankton and heterotrophic prokaryote production positively correlated. Addition of nitrogen alone (+N) had a smaller effect on community production, resulting in metabolically balanced systems (NCP≍0). Finally, heterotrophic conditions persisted in the +P treatment at the central and eastern stations, and gross production was not sufficient to supply bacterial carbon demand, evidence of a decoupling of phytoplankton production and consumption by heterotrophic prokaryotes.

  13. Enhancing a safe water intervention with student-created visual aids to promote handwashing behavior in Kenyan primary schools.

    Science.gov (United States)

    Graves, Janessa M; Daniell, William E; Harris, Julie R; Obure, Alfredo F X O; Quick, Robert

    The Nyando Integrated Child Health Education (NICHE) project was a collaborative effort by the U.S. Centers for Disease Control and local partners to assess the effectiveness of multiple interventions for improving child survival in western Kenya. To increase handwashing in schools, NICHE trained teachers and installed handwashing stations with treated water and soap in 51 primary schools. This cluster-randomized trial evaluated an additional educational strategy (a poster contest themed, "Handwashing with Soap") to improve handwashing behavior in 23 NICHE primary schools. Pupils were engaged in the poster development. Pupil handwashing behavior was observed unobtrusively at baseline and after four months. Intervention schools displayed a significant increase in the number of handwashing stations and proportion of teacher-supervised stations over the study period. No significant between-group differences of intervention in handwashing frequency, soap availability, or visibility of handwashing stations was observed. Despite finding a limited effect beyond the NICHE intervention, the trial appeared to promote sustainability across some measures.

  14. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  15. Water in everyday life and in science classrooms: analysis of discursive interactions and teaching strategies in primary education

    Directory of Open Access Journals (Sweden)

    Andreza Fortini da Silva

    2012-02-01

    Full Text Available This article examines how a primary teacher establishes links between students' initial contributions on the theme ‘water’ and the elements that will make up the teaching approach of this subject in the science classroom. For this purpose, we examine discursive interactions in the first lessons of a teaching sequence, looking for links between events that are being elicited and developed by the teacher with intense participation of the students. We shall also examine the teaching strategies conducted by the teacher, emphasizing the presence of visual resources in text production activities, understanding them as literacy practices in the context of science lessons. To examine the effectiveness of these strategies and mediational resources, we shall analyze some exemplars of the students' productions (texts and drawings. We will use as criteria of analysis: speech marks of the opening activity and of the preliminary discussions in the texts produced by the pupils; evidence of changes in the pupils’ initial repertoires about the theme; evidence of connections between the “water in our lives” and “water as a science subject”. The context of the research is a third year grade classroom in a public elementary school in Contagem / MG - Brazil.

  16. Water in HD 209458b's atmosphere from 3.6 - 8 microns IRAC photometric observations in primary transit

    CERN Document Server

    Beaulieu, J P; Batista, V; Tinetti, G; Ribas, I; Carey, S; Noriega-Crespo, J A; Griffith, C A; Campanella, G; Dong, S; Tennyson, J; Barber, R J; Deroo, P; Fossey, S J; Liang, D; Swain, M R; Yung, Y; Allard, N

    2009-01-01

    The hot Jupiter HD 209458b was observed during primary transit at 3.6, 4.5, 5.8 and 8.0 microns using the Infrared Array Camera (IRAC) on the Spitzer Space Telescope. We detail here the procedures we adopted to correct for the systematic trends present in the IRAC data. The light curves were fitted including limb darkening effects and fitted using Markov Chain Monte Carlo and prayer-bead Monte Carlo techniques, finding almost identical results. The final depth measurements obtained by a combined Markov Chain Monte Carlo fit are at 3.6 microns, 1.469 +- 0.013 % and 1.448 +- 0.013 %; at 4.5 microns, 1.478 +- 0.017 % ; at 5.8 microns, 1.549 +- 0.015 % and at 8.0 microns 1.535 +- 0.011 %. Our results clearly indicate the presence of water in the planetary atmosphere. Our broad band photometric measurements with IRAC prevent us from determining the additional presence of other other molecules such as CO, CO2 and methane for which spectroscopy is needed. While water vapour with a mixing ratio of 10^-4-10^-3 combine...

  17. Carbon dioxide flux and net primary production of a boreal treed bog: responses to warming and water table manipulations

    Science.gov (United States)

    Munir, T. M.; Perkins, M.; Kaing, E.; Strack, M.

    2014-09-01

    Mid-latitude treed bogs are significant carbon (C) stocks and are highly sensitive to global climate change. In a dry continental treed bog, we compared three sites; control, recent (1-3 years; experimental) and older drained (10-13 years; drained) with water levels at 38, 74 and 120 cm below the surface, respectively. At each site we measured carbon dioxide (CO2) fluxes and tree root respiration (Rr) (across hummock-hollow microtopography of the forest floor) and net primary production (NPP) of trees during the growing seasons (May to October) of 2011-2013. The carbon (C) balance was calculated by adding net CO2 exchange of the forest floor (NEff-Rr) to the NPP of the trees. From cooler and wetter 2011 to driest and warmest 2013, The control site was a~C sink of 92, 70 and 76 g m-2, experimental site was a C source of 14, 57 and 135 g m-2, and drained site was a progressively smaller source of 26, 23 and 13 g m-2, respectively. Although all microforms at the experimental site had large net CO2 emissions, the longer-term drainage and deeper water level at the drained site resulted in the replacement of mosses with vascular plants (shrubs) at the hummocks and lichens at the hollows leading to the highest CO2 uptake at drained hummocks and significant losses at hollows. The tree NPP was highest at the drained site. We also quantified the impact of climatic warming at all water table treatments by equipping additional plots with open-top chambers (OTCs) that caused a passive warming on average of ∼1 °C and differential air warming of ∼6 °C (at mid-day full sun) across the study years. Warming significantly enhanced the shrub growth and CO2 sink function of the drained hummocks (exceeding the cumulative respiration losses at hollows induced by the lowered water level × warming). There was an interaction of water level with warming across hummocks that resulted in largest net CO2 uptake at warmed drained hummocks. Thus in 2013, the warming treatment enhanced the

  18. A Novel Computerized Water Level Control System of PWR Steam Generator of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    1998-01-01

    NomenclaturesAArea(m2)EEnergyControl(J)fPressureLossCoeficientaVoidFractionHSpecificEnthaply(J/kg)IInertance(m)KFrictionLossC...

  19. Theoretical estimation of the impact velocity during the PWR spent drop in water condition

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Lee, Seong Ki; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2016-06-15

    The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the 3×3 short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

  20. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  1. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  2. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  3. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  4. 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.

  5. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    CERN Document Server

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  6. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  7. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  8. 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)

  9. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  10. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  11. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2015-10-15

    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  12. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  13. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  14. 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.

  15. The pluralistic water research concept - a new human-water system research approach

    Science.gov (United States)

    Evers, Mariele; Höllermann, Britta; Almoradie, Adrian; Taft, Linda; Garcia-Santos, Glenda

    2017-04-01

    Sustainable water resources management has been and still is a main challenge for decision makers even though for the past number of decades integrative approaches and concepts (e.g. Integrated Water Resources Management - IWRM) have been developed to address problems on floods, droughts, water quality, water quantity, environment and ecology. Although somehow these approaches are aiming to address water related problems in an integrative approach and to some extent include or involve society in the planning and management, they still lack some of the vital components in including the social dimensions and their interaction with water. Understanding these dynamics in a holistic way and how they are shaped by time and space may tackle these shortcomings and provide more effective and sustainable management solutions with respect to a set of potential present social actions and values as well as possible futures. This paper aims to discuss challenges to coherently and comprehensively integrate the social dimensions of different human-water concepts like IWRM, socio-hydrology and waterscape. Against this background it will develop criteria for an integrative approach and present a newly developed concept termed pluralistic water research (PWR) concept. PWR is not only a pluralistic but also an integrative and interdisciplinary approach to acknowledge the social and water dimensions and their interaction and dynamics by considering more than one perspective of a water-related issue, hereby providing a set of multiple (future) developments. Our PWR concept will be illustrated by a case study application of the Canary island La Gomera. Furthermore an outlook on further possible developments of the PWR concept will be presented and discussed.

  16. Recyclable Polymer-Supported Terpyridine–Palladium Complex for the Tandem Aminocarbonylation of Aryl Iodides to Primary Amides in Water Using NaN3 as Ammonia Equivalent

    Directory of Open Access Journals (Sweden)

    Toshimasa Suzuka

    2017-04-01

    Full Text Available Primary aromatic amides are valuable compounds, which are generally prepared via Beckmann rearrangement of oximes and the hydration of nitriles in organic solvents. We investigated the environmentally friendly catalytic aminocarbonylation in water. Thus, a novel heterogeneous transition-metal catalyst, a polymer-supported terpyridine–palladium(II complex, was prepared and found to promote azidocarbonylation of aryl iodides with NaN3 and to reduce the generated benzoyl azides in water under CO gas to yield primary aryl amides with high to excellent yield in a one-pot reaction. The catalyst was recovered and reused several times with no loss of catalytic activity.

  17. Evaluation of SG blowdown demineralizer performance by replacement of ammonia with ethanolamine as a PWR secondary pH control agent

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H. [Department of Materials and Chemical Engineering, Soonchunhyang Univ. (Korea, Republic of); Yim, S.J. [Operation Management Team, Korea Hydro and Nuclear Power Co. Ltd., Seoul (Korea, Republic of)

    2002-07-01

    Four Korean PWR plants have adopted ethanolamine (ETA) as a secondary pH control agent to increase the pH at the liquid phase, which raises the pH in the SG blowdown system. The run time of the SG blowdown demineralizer can be reduced by the increased number of ionic chemical species primarily due to ETA. Contrary to the possible prevention of SG degradation, the replacement of ammonia with ETA results in the water chemistry difficulties and more frequent generation of spent resin. A comprehensive experimental data set for binary, ternary, quaternary, and quinary cation and anion adsorption was developed from small-volume batch tests at total cation or anion concentrations of 0.01 and 0.05 N to obtain the selectivity coefficients of many cations and anions normally present in the PWR secondary system water. In addition, the kinetic study using the bench-scale column was performed to examine the breakthrough point of an ion and to calculate the ratio of inlet to outlet concentration at the column, so called Decontamination Factor, in the different background electrolyte. The batch equilibrium tests indicated that the ion selectivity is higher for an ion of higher valence and is not uniform in the different composition and ionic strength. The preference of an ion on ion exchange resin rather tends to be lower with higher ionic strength. The leakage of an ion from the ion exchange column is not also uniform in time in the various composition and total concentration. Therefore the ion selectivity and breakthrough time are different in ammonia and ethanolamine background electrolytes. The run time of SG blowdown demineralizer can be shorter than it can be expected due to the elevated ionic strength as well as the increased dissolved solids. The quantitative run time can be estimated by such ion exchange models as semi-empirical mass action and surface complexation models. The demineralizer can be used longer by increasing the ratios of cation to anion exchange resins in

  18. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  19. Determination of primary aromatic amines in water food simulant using solid-phase analytical derivatization followed by gas chromatography coupled with mass spectrometry.

    Science.gov (United States)

    Brede, Cato; Skjevrak, Ingun; Herikstad, Hallgeir

    2003-01-03

    Solid phase analytical derivatization with trifluoroacetic anhydride has been introduced as sample preparation for the determination of primary aromatic amines in water by gas chromatography coupled with mass spectrometry. Water was used as a food simulant for testing migration from laminated flexible food packaging materials. The method was evaluated for 8 primary aromatic amines in 200 ml water samples, which resulted in detection limits in the 0.1-0.4 microg/l range, relative standard deviations in the 4-17% range and acceptable linearity (R2 = 0.997-1.000). Detectable levels of 2,4-diaminotoluene, 2,6-diaminotoluene and 4,4'-methylenedianiline were found in water food simulant from some of the investigated food packaging materials.

  20. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki

    1982-06-01

    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  1. Studies of air showers produced by primaries 10(16) eV using a combined scintillation and water-Cerenkov array

    Science.gov (United States)

    Brooke, G.; Perrett, J. C.; Watson, A. A.

    1986-01-01

    An array of 8 x 1.0 sq m plastic scintillation counters and 13 water-Cerenkov detectors (1 to 13.5 sq m) were operated at the center of the Haverah Park array to study some features of air showers produced by 10(16) eV primaries. Measurements of the scintillator lateral distribution function, the water-Cerenkov lateral distribution function, and of the distance dependence of the Cerenkov/scintillator ratio are described.

  2. Experiment data report for LOFT nonnuclear Test L1-4. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Batt, D. L.

    1977-07-01

    Test L1-4 was the fourth in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. Test L1-4 was the first Nuclear Regulatory Commission standard problem (International Problem No. 5 and U.S. Problem No. 7) experiment conducted at LOFT. Data from this test will be compared with predictions generated by the standard problem participants. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200% double-ended offset shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were temperature at 279/sup 0/C, gauge pressure at 15.65 MPa, and intact loop flow at 268.4 kg/s. During system depressurization into a simulated containment, emergency core cooling water was injected into the primary coolant system cold leg to provide data on the effects of emergency core cooling on system thermalhydraulic response.

  3. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  4. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  5. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  6. Corrosion Control of Alloy 690 by Shot Peening and Electropolishing under Simulated Primary Water Condition of PWRs

    Directory of Open Access Journals (Sweden)

    Kyung Mo Kim

    2015-01-01

    Full Text Available This work clarifies the effect of surface modifications on the corrosion rate of Alloy 690, a nickel-based alloy for steam generator tubes, under the simulated test conditions of the primary water chemistry in nuclear power plants. The surface stress was modified by the shot peening and electropolishing methods. The shot peening treatment was applied using ceramic beads with different intensities by varying the air pressure and projection angle. The corrosion rate was evaluated by gravimetric analysis and the surface was analyzed by scanning electron microscopy (SEM. The corrosion rate of Alloy 690 was evaluated from the influence of the stress state on the metal surface. Based on the observation of the surface after the corrosion test, the oxide composition and its structure were affected by the surface modifications. The corrosion behavior of Alloy 690 was distinguished by the shot peening intensity on the surface, and additional electropolishing was effective at reducing the dissolution of nickel ions from the metal surface.

  7. Catalytic supercritical water gasification of primary paper sludge using a homogeneous and heterogeneous catalyst: Experimental vs thermodynamic equilibrium results.

    Science.gov (United States)

    Louw, Jeanne; Schwarz, Cara E; Burger, Andries J

    2016-02-01

    H2, CH4, CO and CO2 yields were measured during supercritical water gasification (SCWG) of primary paper waste sludge (PWS) at 450°C. Comparing these yields with calculated thermodynamic equilibrium values offer an improved understanding of conditions required to produce near-equilibrium yields. Experiments were conducted at different catalyst loads (0-1g/gPWS) and different reaction times (15-120min) in a batch reactor, using either K2CO3 or Ni/Al2O3-SiO2 as catalyst. K2CO3 up to 1g/gPWS increased the H2 yield significantly to 7.5mol/kgPWS. However, these yields and composition were far from equilibrium values, with carbon efficiency (CE) and energy recovery (ER) of only 29% and 20%, respectively. Addition of 0.5-1g/gPWS Ni/Al2O3-SiO2 resulted in high H2 and CH4 yields (6.8 and 14.8mol/kgPWS), CE of 84-90%, ER of 83% and a gas composition relatively close to the equilibrium values (at hold times of 60-120min).

  8. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  9. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  10. An Educational Resource on Water and Health as a Teaching Aid in French Primary Schools--Part II: Design and Validation

    Science.gov (United States)

    Savanovitch, Chantal; Sauvant-Rochat, Marie-Pierre

    2013-01-01

    An educational resource on water and health, using an approach focused on health education and environmental health education, was developed to help teachers in the classroom. The implementation of health education programs in French primary schools is explained. Three specific objectives were identified, targeting 3rd, 4th, and 5th year pupils:…

  11. State of the art of water chemistry relating to preventive maintenance in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Shunji; Hirano, Hideo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    2000-07-01

    Water chemistry of the light water reactors (LWRs) is conducted under aims at ensuring integrity of materials composing the primary system, ensuring integrity of fuel rods, suppressing radiation exposure dose on the plant workers, and ensuring volume reduction of the radioactive wastes. A variety of countermeasures, including modification of materials and structures have been taken for the above aims in addition to the water chemistry. Recently, countermeasures for SCC of in-core structures and control of scale deposition in the secondary system of SG are increasingly required as further countermeasures of preventive maintenance of the BWR and the PWR, respectively. As countermeasures from aspect of the water chemistry, some studies are in progress on application of hydrogen injection and noble metal chemical addition along it for in-core SCC of the BWR. Among 51 units of the LWR under operation in Japan at present, four units have been operated for more than 25 years. Studies have been in progress, therefore, on the water chemistry technology of the preventive maintenance for long-term aging. Here was also described on trends in the preventive maintenance water chemistry for the BWR in Japan. (G.K.)

  12. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  13. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the {kappa}-{epsilon} turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and

  14. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  15. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  16. Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-10-01

    This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor (BRPR).

  17. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  18. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  19. 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.

  20. Water

    Science.gov (United States)

    ... www.girlshealth.gov/ Home Nutrition Nutrition basics Water Water Did you know that water makes up more ... to drink more water Other drinks How much water do you need? top Water is very important, ...