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Sample records for transfer region pwr

  1. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  2. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 103

    Energy Technology Data Exchange (ETDEWEB)

    Clemons, V.D.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-03-07

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 103, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system.

  3. 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

  4. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  5. Assessment and limitation of radioactivity transfers in the event of a postulated severe PWR accident

    International Nuclear Information System (INIS)

    Gauvain, J.

    1992-01-01

    This report constitutes the supporting material for a lecture on severe accidents which could occur on PWR type nuclear reactors. It is assumed for present purposes that the reader has at least a rudimentary acquaintance with the basics of general physics if not with the operating processes of these reactors. After defining what is meant by a ''severe accident'' on a reactor, the possible phenomenology of such an accident is qualitatively described: loss of coolant and loss of containment integrity. A certain number of elements are then given for the quantitative assessment of these phenomena involving possible radioactivity transfers within and outside the plant. In conclusion, available means are indicated for the limitation and control of these environmental transfers. (author). 5 refs, figs

  6. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  7. 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)

  8. Study on PCS heat and mass transfer of advanced PWR with CFD code

    Energy Technology Data Exchange (ETDEWEB)

    Huang, X. G.; Cheng, X. [Shanghai Jiao Tong Univ., Shanghai (China); Wang, F. N.; Zhang, Z. D.; Cheng, X. [State Nuclear Power Technology Company, Beijing (China)

    2012-03-15

    During the hypothetical Double-Ended Cold Leg Guillotine (DECLG) of large advanced pressure water reactor (PWR), a large amount of steam ejects from the break into the containment. Passive containment cooling system (PCS) is implemented to prevent over-pressure and over-temperature. The computational fluid dynamics (CFD) code GASFLOW coupled with Film Coverage and Evaporation Model (FICEM) is applied in this study to analyze the PCS performance during DECLG.FICEM can calculate film coverage rate, film evaporation rate and containment heat removal capability. Results show that the modified GASFLOW version coupled with FICEM is feasible to analyze the thermal-hydraulic behavior in PCS of advanced passive PWR. Capability of PCS for large scale PWR is investigated through using the modified GASFLOW code.

  9. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    International Nuclear Information System (INIS)

    1976-02-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation

  10. Effect of heat transfer in the fog region during core reflooding

    International Nuclear Information System (INIS)

    Rouai, N. M.; El-sawy, H. M.

    1993-01-01

    Core reflooding following a loss of coolant accident (LOCA) in a pressurized water reactor (PWR) received considerable attention during the past thirty years. In this paper a one dimensional model is used to study the effect of the heat transfer in the fog region ahead of the wet front reflooding rate of a cylindrical fuel element following a LOCA in a PWR. The heat conduction equation in the cladding is solved in coordinate system moving with the wet front under a variety of condition to investigate the effects of such parameters as the initial cladding surface temperature, the decay heat generation rate in the fuel and the mode of heat transfer prevailing. The cladding surface is divided into three axial regions according to the mechanism of heat transfer, namely, a boiling region behind the wet front, a fog region ahead of the wet front and a dry region further downstream of the wet front. The effect of changing the heat transfer coefficient in the fog region on the rewetting rate and on the fog length is investigated. The results of this simple model show that increasing the heat transfer in the fog region increases the rewetting velocity and consequently decreases the fog length. The results are in general agreement with a more accurate two-dimensional model and experimental data. (author)

  11. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  12. A new correlation for convective heat transfer coefficient of water–alumina nanofluid in a square array subchannel under PWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair A. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Mechanical Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Bhowmik, Palash K. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Department of Nuclear Engineering, Missouri University of Science and Technology, 1201 N. State St., Rolla, MO 65409 (United States); Xiangyi, Chen [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of); Suh, Kune Y., E-mail: kysuh@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

    2016-11-15

    Highlights: • Thermo-hydrodynamic properties of water–Al{sub 2}O{sub 3} nanofluid at PWR condition is analyzed. • Details of CFD simulation and validation procedure is outlined. • Augmented heat transfer capacity of nanofluid is governed by larger pumping power. • A new correlation for nanofluid Nusselt number in subchannel geometry is proposed. - Abstract: The computational fluid dynamic (CFD) simulation is performed to determine on the thermo- and hydrodynamic performance of the water–alumina (Al{sub 2}O{sub 3}) nanofluid in a square array subchannel featuring pitch-to-diameter ratios of 1.25 and 1.35. Two fundamental aspects of thermal hydraulics, viz. heat transfer and pressure drop, are assessed under typical pressurized water reactor (PWR) conditions at various flow rates (3 × 10{sup 5} ⩽ Re ⩽ 6 × 10{sup 5}) using pure water and differing concentrations of water–alumina nanofluid (0.5–3.0 vol.%) as coolant. Numerical results are compared against predictions made by conventional single-phase convective heat transfer and pressure loss correlations for fully developed turbulent flow. It is observed that addition of tiny nanoparticles in PWR coolant can give rise to the convective heat transfer coefficient at the expense of larger pressure drop. Nevertheless, a modified correlation as a function of nanoparticle volume fraction is proposed to estimate nanofluid Nusselt number more precisely in square array subchannel.

  13. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

  14. Modeling of heat transfers inside the bottom of a PWR reactor vessel during a serious accident

    International Nuclear Information System (INIS)

    Simeoni, A.; Tarabelli, D.; Ratel, G.; Spindler, B.; Tourniaire, B.; Guillard, G.

    2006-01-01

    In the framework of its studies on serious reactor accidents, the French atomic energy commission (CEA) is developing models describing the thermal behaviour of the corium (magma made of the molten core with structure materials) at the bottom of the pressure vessel, using results coming from the most recent experimental programs. This knowledge is capitalized with the scenario code ASTEC developed by IRSN and GRS. The article presents the models entered in ASTEC to calculate the heat transfers inside the corium bath and the results in terms of heat flux exchanged with the vessel. (J.S.)

  15. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  16. Achieving regionalization through rural interhospital transfer.

    Science.gov (United States)

    Feazel, Leah; Schlichting, Adam B; Bell, Gregory R; Shane, Dan M; Ahmed, Azeemuddin; Faine, Brett; Nugent, Andrew; Mohr, Nicholas M

    2015-09-01

    Regionalization of emergency medical care aims to provide consistent and efficient high-quality care leading to optimal clinical outcomes by matching patient needs with appropriate resources at a network of hospitals. Regionalized care has been shown to improve outcomes in trauma, myocardial infarction, stroke, cardiac arrest, and acute respiratory distress syndrome. In rural areas, effective regionalization often requires interhospital transfer. The decision to transfer is complex and includes such factors as capabilities of the presenting hospital; capacity at the receiving hospital; and financial, geographic, and patient-preference considerations. Although transfer to a comprehensive center has proven benefits for some conditions, the transfer process is not without risk. These risks include clinical deterioration, limited resource availability during transport, vehicular crashes, time delays for time-sensitive care, poor communication between providers, and neglect of patient preferences. This article reviews the transfer decision, financial implications, risks, and considerations for patients undergoing rural interhospital transfer. We identify several strategies that should be considered for development of the regionalized emergency health care system of the future and identify areas where further research is necessary. Copyright © 2015 Elsevier Inc. All rights reserved.

  17. The one-dimensional normalised generalised equivalence theory (NGET) for generating equivalent diffusion theory group constants for PWR reflector regions

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1991-01-01

    An equivalent diffusion theory PWR reflector model is presented, which has as its basis Smith's generalisation of Koebke's Equivalent Theory. This method is an adaptation, in one-dimensional slab geometry, of the Generalised Equivalence Theory (GET). Since the method involves the renormalisation of the GET discontinuity factors at nodal interfaces, it is called the Normalised Generalised Equivalence Theory (NGET) method. The advantages of the NGET method for modelling the ex-core nodes of a PWR are summarized. 23 refs

  18. An assessment of the failure rate for the beltline region of PWR pressure vessels during normal operation and certain transient conditions. Technical report

    International Nuclear Information System (INIS)

    Gamble, R.M.; Strosnider, J. Jr.

    1981-06-01

    A study was conducted to assess the failure rate for the beltline region of a generic pressurized-water reactor (PWR) pressure vessel. This assessment included the evaluation of several normal operating and transient reactor conditions. Failure rates were calculated from a computer code that used fracture mechanics methods to model the failure process; random number generation techniques were used to simulate random variables and model their interaction in the failure-process. This investigation had three major objectives: (1) to better define the effect of neutron irradiation, material variation, and flaw distribution on the failure rate for the beltline region of PWR pressure vessels, (2) to estimate the relative margins against failure for normal operation and certain transient conditions associated with nuclear pressure vessels, and (3) to evaluate the current limitations for using fracture mechanics models to predict failure rates for nuclear pressure vessels

  19. Evaluation of the heat transfer in a geological repository concept containing PWR, VHTR and hybrid ads-fission spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jonusan, Raoni A.S.; Pereira, Fernando; Velasquez, Carlos E.; Salome, Jean A.D.; Cardoso, Fabiano; Pereira, Claubia; Fortini, Angela, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    The investigation of the thermal behavior of spent fuel (SF) materials is essential to determining appropriate potential sites to accommodate geological repositories as well as the design of canisters, considering their potential risk to people health and of environmental contamination. This work presents studies of the temperature in a canister containing spent fuels discharged from Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) reactor systems in a geological repository concept. The thermal analyses were performed with the software ANSYS, which is widely used to solve engineering problems through the Finite Element Method. The ANSYS Transient Thermal module was used. The spent nuclear fuels were set as heat sources using data of previous studies derived from decay heat curves. The studies were based on comparison of the mean temperature on a canister surface along the time under geological disposal conditions, for a same amount of each type of spent nuclear fuel evaluated. The results conclude that fuels from VHTR and ADS systems are inappropriate to be disposed in a standardized PWR canister, demanding new studies to determine the optimal amount of spent fuel and new internal canister geometries. It is also possible to conclude that the hypothetical situation of a single type of canister being used to accommodate different types of spent nuclear fuels is not technically feasible. (author)

  20. Axial Region Optimization for Cycle Length Extension of Small Modular PWR

    Energy Technology Data Exchange (ETDEWEB)

    Choe, Ji Won [UNIST, Ulsan (Korea, Republic of); Shin, Ho Cheol; Jung, Ji Eun [KHNP CRI, Daejeon (Korea, Republic of); Zheng, Youqi; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    This paper studied axial region sensitivity test in SMPWR to extend the cycle length. Zr- {sup 167}Er, Zr-B and Zr-Gd are used for BA material, the height of cutback and axial region has been changed. The cycle length of the reference core is 798 EFPD (26.6 EFPM), and there is no cutback and only {sup 167}Er-Zr is used in R-BA.Soluble boron-free small modular pressurized water reactor (SMPWR) can be a transportable size core due to the absence of the chemical volume control system (CVCS) and the amount of liquid radioactive waste, and further remove the corrosion problem caused by boric acid. The SMPWR needs large amount of burnable absorber (BA) instead of soluble boron, but the more the amount of BA is, the shorter the fuel cycle length is. Therefore, this paper studies axial region sensitivity test to make fuel cycle length of SMWPR longer. The procedure of axial region sensitivity test is as follows: cutback sensitivity tests, material and height sensitivity tests in the axial's reactor core design code system, has been used for these simulation. The optimal BA for cutback region is 10 cm of cutback with natural Gd 10 % in Zr-Gd, and the cycle length increases to 942 EFPD (31.4 months). Through the axial region sensitivity test, the cycle length becomes 1026 EFPD (34.2 months), but the peaking factors were not satisfied their limits. The 4.8-month increases compared with the reference core through the cutback sensitivity test. The possibility to excess reactivity with control rods in this core should verify.

  1. The Regional Test Center Data Transfer System

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Daniel M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Photovoltaic and Distributed Systems Dept.; Stein, Joshua S. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Photovoltaic and Distributed Systems Dept.

    2016-09-01

    The Regional Test Centers are a group of several sites around the US for testing photovoltaic systems and components related to photovoltaic systems. The RTCs are managed by Sandia National Laboratories. The data collected by the RTCs must be transmitted to Sandia for storage, analysis, and reporting. This document describes the methods that transfer the data between remote sites and Sandia as well as data movement within Sandia’s network. The methods described are in force as of September, 2016.

  2. Estimation of droplets/wall heat transfer under LOCA conditions in a PWR; Estimation du transfert de chaleur gouttes/paroi en situation d'APRP pour un REP

    Energy Technology Data Exchange (ETDEWEB)

    GrAdeck, M.; Maillet, D. [CNRS UMR 7563 2, 54 - Vandoeuvre les Nancy (France); Lelong, F.; Seiler, N.; Repetto, G. [IRSN Cadarache, 13 - Saint Paul lez Durance (France)

    2009-07-01

    During a LOCA (Loss Of Coolant Accident) in a PWR, the fuel assemblies could be locally severely ballooned. The transient is ended by the injection of water initiated the safety system. The cooling of theses partially blocked fuel assemblies depends on the coolant flow characteristics in the blockage region. Most models for heat transfers concentrate on cooling of the ballooned walls by vapor convection. Since a two-phase mist flow occurs when reflooding, the possibility of additional cooling by direct liquid droplet impingement on the blockage surfaces must be investigated. As the temperature of the fuel assemblies is higher than the Leidenfrost temperature, the impact regime should be only the bouncing one. Up to now, no model of heat transfer of droplet impacts has been developed for that regime. As the coolability from droplet impacts must be modeled, an experimental program was proposed with droplets and wall characteristics (velocity, diameter, temperature) close to the LOCA ones. As the interaction between the droplet and the wall is very short (a few of ms), the estimation of the heat flux during the resident time of the droplet at the wall must be accurately designed. The purpose of this work is to show how such heat flux can be experimentally estimated used an adapted inverse heat conduction model. The final goal of the present collaboration between LEMTA and IRSN is to introduce the cooling model within NEPTUNE-CFD code, a joint project of CEA, EDF, AREVA and IRSN. (authors)

  3. Estimation of heat transfer rates to droplets under the conditions of a LOCA in a PWR in the ballooned zone

    Energy Technology Data Exchange (ETDEWEB)

    Gradeck, Michel; Maillet, Denis [LEMTA Nancy-University CNRS, 2 av de la foret de Haye, BP160, 54504 Vandoeuvre cedex (France); Lelong, Franck [LEMTA Nancy-University CNRS, 2 av de la foret de Haye, BP160, 54504 Vandoeuvre cedex (France)]|[IRSN/DPAM/SEMCA/LEIDC, Cadarache Batiment 700, BP3 - 13 115 Saint Paul lez Durance cedex (France); Seiler, Nathalie [IRSN/DPAM/SEMCA/LEIDC, Cadarache Batiment 700, BP3 - 13 115 Saint Paul lez Durance cedex (France)

    2008-07-01

    Full text of publication follows: During a LOCA (Loss Of Coolant Accident), the critical regions (in terms of safety) of the fuel assemblies could be ballooned. The cooling of theses partially blocked fuel assemblies depends on the coolant flow characteristics in the blockage region. Most models for heat transfer concentrate on cooling of the ballooned walls by vapor convection. Since a two-phase mist flow occurs when reflooding, the possibility of additional cooling by direct liquid droplet impingement on the blockage surfaces must be investigated. As the temperature of the fuel assemblies is higher than the Leidenfrost temperature, the impact regime should be only the bouncing one. Up to now, no model of heat transfer of droplet impacts has been developed for that regime. As the coolability from droplet impacts must be modeled, we realize an experimental study with droplets and wall characteristics (velocity, diameter, temperature) close to the LOCA ones. As the interaction between the droplet and the wall is very short (a few of ms), the estimation of the heat flux during the resident time of the droplet at the wall must be accurately designed. The purpose of this work is to show how such heat flux can be experimentally estimated used an adapted inverse heat conduction model. The final goal of the present collaboration between LEMTA (Laboratory of Applied and Theoretical Energy and Mechanic) and IRSN (Institut of Radioprotection and Nuclear Safety) is to introduce the cooling model within NEPTUNE-CFD code of the NEPTUNE thermal-hydraulic platform, a joint project of CEA, EDF, IRSN and AREVA. (authors) [French] Dans le cas d'un APRP (Accident de Perte de Refrigerant Primaire), les zones critiques de l'assemblage combustible peuvent etre deformees. Le refroidissement de ces zones depend de l'importance du blocage qui affectera l'ecoulement diphasique les traversant. La plupart des modeles de refroidissement de ces zones assechees, a hautes

  4. Droplets flow and heat transfer at top region of core in reflood phase

    International Nuclear Information System (INIS)

    Osakabe, Masahiro; Ohnuki, Akira; Sobajima, Makoto

    1983-02-01

    The heat transfer at the top region of core is complicated due to the strong thermal non-equilibrium just after the start of reflood phase in a postulated PWR-LOCA experiment. The film taken with a high-speed cinecamera shows upward droplets flow and falling water film on the non-heated rod just after the start of reflood at elevation 3235 mm above the bottom of heated length of heater rods in Slab Core Reflood Test. The measured mean diameter of droplet is about 1 mm. This value of mean diameter is larger than the measured result for the annular dispersed flow in a pipe. On the other hand, the corresponding Weber number is smaller than the Weber number in the accelerating flow obtained in the previous studies. The calculated heat transfer coefficient of the droplets flow approximately agrees with the sum of Dittus-Boelter's convective heat transfer term and radiative heat transfer term evaluated with the network analysis by Sun et al. (author)

  5. Analyses of rewetting in water reactor emergency core cooling inclusive of heat transfer in the unwetted region

    International Nuclear Information System (INIS)

    Chun, M.H.

    1976-01-01

    One-dimensional analysis of rewetting a vertical hot surface is carried out without neglecting heat transfer in the unwetted region. The physical model consists of an infinitely extended vertical thin slab whose initial surface temperature is higher than the Leidenfrost point. During rewetting, three different regions of constant heat transfer coefficients are assumed: (1) the surface of the wetted region as characterized by a higher constant heat transfer coefficient (h/sub c/), (2) the two regions of unwetted surface which is precooled via a lower constant heat transfer coefficient (h/sub df/) for the fog-film region, and (3) the negligible heat transfer coefficient for the superheated vapor-film region (h/sub ds/). An explicit formula which computes the wet front velocity is obtained for the more general case when the value of h/sub df/ is a finite value. A quantitative analyses of phi and fog length (l), which denotes the fraction of unevaporated water droplets (and vapor) constituting fog-film flow and the distance between the wet and fog front, respectively, are carried out using the most typical bottom-flooding experimental results as reported by Case et al., and PWR-FLECHT report. The results show that phi values fall within the upper and lower bounds derived from theory

  6. Modelling local chemistry in PWR steam generator crevices

    International Nuclear Information System (INIS)

    The accumulation of impurities in local regions of PWR Steam Generators (SG) has resulted in the accelerated corrosion of SG materials. The chemical conditions in crevices and sludge piles is dependent on thermal hydraulic and mass transfer processes as well as the physical chemistry of the concentrated solution itself. This paper discusses the different modelling approaches which can be used to describe the concentration process and the local chemistry in these regions. The limitations of each approach and the applicability of model results to field conditions are discussed in the paper. EPRI's program in this area, including past accomplishments and the models used in the MULTEQ code are described in the paper. (author)

  7. Simulation of the stationary and unstationary thermohydraulic processes in PWR reactors using RELAP 3 with the assumption of a forced lateral exchange between highly stressed core regions and core regions under normal stress

    International Nuclear Information System (INIS)

    Welhusen, B.

    1976-03-01

    LOCASs in LWR reactors are studied at the IKE using the code RELAP 3. In earlier calculations with RELAP 3, only the axial power density distribution was taken into account in determining core regions for code input, i.e. the core consisted of several control volumes connected in series. In the present paper, the influence of axial and radial power density distribution in the core on the stationary and instationary behaviour of the core of a PWR reactor is studied using RELAP 3. (orig./TK) [de

  8. Heat transfer to liquid sodium in the thermal entrance region

    International Nuclear Information System (INIS)

    Qiu, R.

    1981-01-01

    It is well known that the convective heat transfer in the regions of duct systems where the thermal boundary layers are not yet established can be far superior to heat transfer in the fully developed regions. A quantitative understanding of heat transfer in the thermal entrance region is essential in designing high heat-flux nuclear reactors. More specifically, if the thermal boundary layers have not been fully established in the system, the forced-convection relations for the fully developed regions cannot be used to predict the heat transfer characteristics. The present work is characterized by the following: 1. The behaviours in the thermal entrance region have been examined more completely. 2. To obtain a higher accuracy of analyses, in present study the method of SPARROW et al. for pipe was improved for annulus by utilizing a finite difference technique. Furthermore, an asymptotic solution was developed. 3. This is, in our knowledge, the first experimental investigation about the thermal development effect on turbulent heat transfer from rod element to liquid sodium in annulus with fully developed flow. (MDC)

  9. Stagnation Region Heat Transfer Augmentation at Very High Turbulence Levels

    Energy Technology Data Exchange (ETDEWEB)

    Ames, Forrest [University of North Dakota; Kingery, Joseph E. [University of North Dakota

    2015-06-17

    A database for stagnation region heat transfer has been extended to include heat transfer measurements acquired downstream from a new high intensity turbulence generator. This work was motivated by gas turbine industry heat transfer designers who deal with heat transfer environments with increasing Reynolds numbers and very high turbulence levels. The new mock aero-combustor turbulence generator produces turbulence levels which average 17.4%, which is 37% higher than the older turbulence generator. The increased level of turbulence is caused by the reduced contraction ratio from the liner to the exit. Heat transfer measurements were acquired on two large cylindrical leading edge test surfaces having a four to one range in leading edge diameter (40.64 cm and 10.16 cm). Gandvarapu and Ames [1] previously acquired heat transfer measurements for six turbulence conditions including three grid conditions, two lower turbulence aero-combustor conditions, and a low turbulence condition. The data are documented and tabulated for an eight to one range in Reynolds numbers for each test surface with Reynolds numbers ranging from 62,500 to 500,000 for the large leading edge and 15,625 to 125,000 for the smaller leading edge. The data show augmentation levels of up to 136% in the stagnation region for the large leading edge. This heat transfer rate is an increase over the previous aero-combustor turbulence generator which had augmentation levels up to 110%. Note, the rate of increase in heat transfer augmentation decreases for the large cylindrical leading edge inferring only a limited level of turbulence intensification in the stagnation region. The smaller cylindrical leading edge shows more consistency with earlier stagnation region heat transfer results correlated on the TRL (Turbulence, Reynolds number, Length scale) parameter. The downstream regions of both test surfaces continue to accelerate the flow but at a much lower rate than the leading edge. Bypass transition occurs

  10. MODEL OF REGIONAL KNOWLEDGE TRANSFER: MAIN ACTORS, FRAMEWORK AND THEORY.

    Directory of Open Access Journals (Sweden)

    Alla LEVITSKAIA

    2016-02-01

    Full Text Available This paper analyses potential mechanism of regional knowledge transfer in region with poorly developed innovation infrastructure (the Autonomous Territorial Unit Gagauzia, Republic of Moldova through interactions between regional major players of the Regional Innovation System - the educational and research institutions, small and medium enterprises (SMEs and local authorities. Solution of this problem can be found in modern studies of territories innovation development through the clustering processes. Through the empirical study - innovation potential analysis of local SMEs - we proposed advantage mechanism which focused on the one type of knowledge cluster – Innovation and Educational Cluster. The symbiosis of entrepreneurs, government agencies, educational institutions and business service providers with the regional core - University, allows to increasing exchange flows of innovative knowledge between all members of the cluster and distributing them to the entire region and beyond. The results and proposals of this study formed the basis of the “Program of increasing the innovation potential of Gagauz SMEs”.

  11. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  12. Transfer Frequency as a Measure of Hospital Capability and Regionalization.

    Science.gov (United States)

    França, Urbano L; McManus, Michael L

    2017-12-01

    To provide metrics for quantifying the capability of hospitals and the degree of care regionalization. Administrative database covering more than 10 million hospital encounters during a 3-year period (2012-2014) in Massachusetts. We calculated the condition-specific probabilities of transfer for all acute care hospitals in Massachusetts and devised two new metrics, the Hospital Capability Index (HCI) and the Regionalization Index (RI), for analyzing hospital systems. The HCI had face validity, accurately differentiating academic, teaching, and community hospitals of varying size. Individual hospital capabilities were clearly revealed in "fingerprints" of their condition-specific transfer behavior. The RI also performed well, with those of specific conditions successfully quantifying the concentration of care arising from regulatory and public health activity. The median RI of all conditions within the Massachusetts health care system was 0.21 (IQR, 0.13-0.36), with a long tail of conditions that were very highly regionalized. Application of the HCI and RI metrics together across the entire state identified the degree of interdependence among its hospitals. Condition-specific transfer activity, as captured in the HCI and RI, provides quantitative measures of hospital capability and regionalization of care. © Health Research and Educational Trust.

  13. Seawater desalination using reusable type small PWR

    Energy Technology Data Exchange (ETDEWEB)

    Uchiyama, Y. [Institute of Engineering Mechanics and Systems, University of Tsukuba, Tsukuba, Ibaraki (Japan); Minato, A. [Planning Division, Central Research Institute of the Electric Power Industry, Komae-shi, Tokyo (Japan); Shimamura, K. [Nuclear Systems Engineering Department, Nuclear Energy Systems Engineering Center, Mitsubishi Heavy Industries, Ltd., Kanagawa (Japan)]. E-mail: shimamura@atom.hq.mhi.co.jp

    2003-07-01

    Demand for seawater desalination is increasing, especially in regions such as the Middle East and North Africa, where populations are growing at a high annual rate. If such demand is met by fossil fuel energy, the influence on the environment, such as global warming, cannot be disregarded. Since these regions are behind in their preparedness of social capital infrastructure, such as power transfer grids, small reactors are considered to be more suitable for introduction than the large reactors found commonly in developed countries. Therefore, a small reusable PWR with mid-range pressure and temperature services, which does not require on-site refuelling, was devised for seawater desalination. In a small reusable PWR, spent fuel is taken out together with the reactor vessel and refuelled on the exterior fuel exchange base prepared independently. Thus, the safeguards against nuclear proliferation increase at a plant site because the lid of the reactor vessel is never opened at the site, in principle. The reactor vessel will be transported from the plant site to a fuel exchange base under stipulated conditions within a transportation cask after a long (about six years) operation. Since fuel handling facilities at the site become unnecessary through centralisation at a fuel exchange base, initial plant construction costs are reduced. In addition, the reactor vessel is reused until its service life has expired. This examination was based on the marine reactor of the experimental nuclear ship, Mutsu, after it had been applied for land use: at a lowered, midrange pressure and temperature service, in theory. It is possible to produce fresh water through reverse osmosis (RO) membrane pressure-rising seawater by a steam turbine driven pump. Using the method of driving a desalination unit high-pressure pump directly by low-pressure steam generated from the heating reactor, fresh water can be produced efficiently. Furthermore, operating at reduced pressure makes it possible

  14. Site specific transfer factor studies for Kaiga region

    International Nuclear Information System (INIS)

    Karunakara, N.

    2012-01-01

    The Radioecology Laboratory of University Science Instrumentation Centre, Mangalore University is engaged in frontline research studies on different aspects of environmental radioactivity and radiation protection for the last 20 years. Extensive studies have been carried out on radiation levels, radionuclides distribution, and transfer of radionuclides through terrestrial, aquatic and atmospheric pathways in the environment of West Coast of India including the Kaiga nuclear power plant. The baseline studies on radioactivity levels around Kaiga region was carried out well before the nuclear power plant became operational and the data generated under these studies are considered to be highly valuable for future impact assessments. The nuclear power plant became operational in the year 1999 and since then this laboratory is involved in radiological impact assessment studies around the nuclear power plant. Detailed Kaiga specific studies are now ongoing to estimate the transfer factors and transfer coefficients for radionuclides for different pathways, such as, (i) soil to rice (ii) soil to different types of vegetables (iii) water/sediment to fish (iv) soil to grass (v) grass to cow milk and (vi) milk to child. For these studies, rice and vegetable fields were developed very close to the nuclear power plant in Kaiga to study the transfer of radionuclides. The water required for this field was drawn from coolant water discharge canal of the power plant. Rice and different types of vegetables were grown in the experimental fields in different seasons of the year and the uptake of radionuclides was studied. For a comparative study, rice and vegetables were also collected from the fields of farmers of nearby villages and analysed. The transfer of artificial radionuclides through pathway involving cow milk was also studied in detail. A grass field was developed and cows were adopted specifically for this study. The cows were allowed to graze freely in this grass field

  15. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    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.)

  16. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  17. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  18. Evaluation of the radiative transfer in the core of a Pressurized Water Reactor (PWR) during the reflooding step of a Loss Of Coolant Accident (LOCA)

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

    We developed a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected. (author)

  19. Challenges of model transferability to data-scarce regions (Invited)

    Science.gov (United States)

    Samaniego, L. E.

    2013-12-01

    Developing the ability to globally predict the movement of water on the land surface at spatial scales from 1 to 5 km constitute one of grand challenges in land surface modelling. Copying with this grand challenge implies that land surface models (LSM) should be able to make reliable predictions across locations and/or scales other than those used for parameter estimation. In addition to that, data scarcity and quality impose further difficulties in attaining reliable predictions of water and energy fluxes at the scales of interest. Current computational limitations impose also seriously limitations to exhaustively investigate the parameter space of LSM over large domains (e.g. greater than half a million square kilometers). Addressing these challenges require holistic approaches that integrate the best techniques available for parameter estimation, field measurements and remotely sensed data at their native resolutions. An attempt to systematically address these issues is the multiscale parameterisation technique (MPR) that links high resolution land surface characteristics with effective model parameters. This technique requires a number of pedo-transfer functions and a much fewer global parameters (i.e. coefficients) to be inferred by calibration in gauged basins. The key advantage of this technique is the quasi-scale independence of the global parameters which enables to estimate global parameters at coarser spatial resolutions and then to transfer them to (ungauged) areas and scales of interest. In this study we show the ability of this technique to reproduce the observed water fluxes and states over a wide range of climate and land surface conditions ranging from humid to semiarid and from sparse to dense forested regions. Results of transferability of global model parameters in space (from humid to semi-arid basins) and across scales (from coarser to finer) clearly indicate the robustness of this technique. Simulations with coarse data sets (e.g. EOBS

  20. University, Knowledge and Regional Development: Factors Affecting Knowledge Transfer in a Developing Region

    Science.gov (United States)

    Fongwa, Neba Samuel; Marais, Lochner

    2016-01-01

    The role of knowledge in the current knowledge economy cannot be overly emphasised. Successful regions are continuously being linked to excellence in the production, accumulation, and application of knowledge. Universities have increasingly been at the centre of such knowledge production, application and transfer. Yet, there is little research and…

  1. Assessment of asynchronous transfer mode (ATM) networks for regional teleradiology

    Science.gov (United States)

    Duerinckx, Andre J.; Hayrapetian, Alek S.; Valentino, Daniel J.; Grant, Edward G.; Rahbar, Darius; Kiszonas, Mike; Franco, Ricky; Shimabuku, Guy H.; Hagan, Girish T.; Melany, Michelle; Narin, Sherelle L.; Ragavendra, Nagesh

    1996-05-01

    The purpose of this study was to assess the effect of ATM network capabilities on the clinical practice of regional teleradiology, by providing immediate interactive radiology consultations between subspecialists and general radiologists at affiliated academic institutions. PACS installed at three affiliated hospitals (UCLA Medical Center, West LA VAMC and UCLA Olive-View Medical Centers) were connected via an ATM network. Two commercial PACS (Agfa) systems, one at the VAMC and one in an ultrasound outpatient clinic at UCLA were connected via ATM switches (Newbridge, Inc.) and a Santa Monica GTE central office switch. We evaluated this initial system configuration and measured image transfer performance, including memory-to-memory, disk-to-disk, disk-to-archive with and without DICOM protocols. Although the memory-to-memory data rate was 25 Mbps, the average remote disk-to-disk image transfer performance, using DICOM 3.0 communications protocols on SUN SPARCstation 10 servers, was 3 to 5 Mbps. Using these capabilities, timely interactive subspecialty consultations between radiologists was successfully performed while both were at different physical locations. We present the use of ATM technology in a realistic clinical environment and evaluate its impact on patient care and clinical teaching within the radiology departments of 2 institutions. Image communications over a regional PACS using an ATM network can allow interactive consultations between different subspecialist and general radiologists or other specialized radiologist spread over three different medical centers.

  2. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    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

  3. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  4. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  5. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  6. TESTING SOME PEDO-TRANSFER FUNCTIONS (PTFS IN APULIA REGION

    Directory of Open Access Journals (Sweden)

    Floriano Buccigrossi

    2009-03-01

    Full Text Available The knowledge of soil water retention vs. soil water matric potential is used to study irrigation and drainage schedules, soil water storage capacity (plant available water, solute movement, plant growth and water stress. The hydraulic soil properties measuring is expensive, laborious and takes too long time, so, frequently, matemathic models, called pedo-transfer functions (PTFs are utilized to estimate hydraulic soil properties through soil chimical and phisical characteristics. Six pedo-transfer functions have been evaluated (Gupta & Larson, 1979; Rawls et al., 1982; De Jong et al., 1983; Rawls & Brakensiek, 1985; Saxton et al., 1986; Vereecken et al., 1989 by comparing estimated with measured soil moisture values at soil water matric potential of –33 and –1500 kPa of 361 soil samples collected from 185 pedons of Apulia Region (South Italy, having various combinations of particle-size distribution, soil organic matter content and bulk density. Accuracy of the soil moisture predictions have been evaluated by statistic indexes such as Weighted stantard error (WSEE, Mean Deviation (MD, Root Mean Squared Deviation (RMSD and the determination coefficient (R2 between estimated and measured water retention values. The Rawls PTF model demostrated to have the lowest values of WSEE, MD and RMSD indexes (0.044, -0.007 and 0.059 m3 H2O m-3 soil, respectively at –33 Kpa soil water matric potential (Field Capacity, while for estimating soil moisture at the Wilting Point (-1500 kPa Rawls & Brakensiek model is adequate (WSEE, MD and RMSD of 0.034, -0.016 and 0.046 m3 H2O m-3 soil. De Jong, Saxton and Rawls & Brakensiek models, at –33 kPa soil water matric potential and Gupta & Larson and De Jong models at –1500 kPa soil water matric potential, showed the highest statistic errors.

  7. Centers tehnology transfer-active factor an the regional development

    OpenAIRE

    Ghimisi, Stefan/St; Popescu, Gheorghe

    2009-01-01

    The purpose of this paper is to investigate mechanisms of knowledge transfer between firms and universities. Universities have become increasingly involved in technology transfer by establishing offices of technology transfer, business incubators, and technology parks. This paper presents some aspects of technology transfer centers, specific activities in these entities, with a real example, UCB-Pitt, an entity founded the University Constantin Brancusi of Targu Jiu.

  8. Transfer of Satellite Rainfall Uncertainty from Gauged to Ungauged Regions at Regional and Seasonal Timescales

    Science.gov (United States)

    Tang, Ling; Hossain, Faisal; Huffman, George J.

    2010-01-01

    Hydrologists and other users need to know the uncertainty of the satellite rainfall data sets across the range of time/space scales over the whole domain of the data set. Here, uncertainty' refers to the general concept of the deviation' of an estimate from the reference (or ground truth) where the deviation may be defined in multiple ways. This uncertainty information can provide insight to the user on the realistic limits of utility, such as hydrologic predictability, that can be achieved with these satellite rainfall data sets. However, satellite rainfall uncertainty estimation requires ground validation (GV) precipitation data. On the other hand, satellite data will be most useful over regions that lack GV data, for example developing countries. This paper addresses the open issues for developing an appropriate uncertainty transfer scheme that can routinely estimate various uncertainty metrics across the globe by leveraging a combination of spatially-dense GV data and temporally sparse surrogate (or proxy) GV data, such as the Tropical Rainfall Measuring Mission (TRMM) Precipitation Radar and the Global Precipitation Measurement (GPM) mission Dual-Frequency Precipitation Radar. The TRMM Multi-satellite Precipitation Analysis (TMPA) products over the US spanning a record of 6 years are used as a representative example of satellite rainfall. It is shown that there exists a quantifiable spatial structure in the uncertainty of satellite data for spatial interpolation. Probabilistic analysis of sampling offered by the existing constellation of passive microwave sensors indicate that transfer of uncertainty for hydrologic applications may be effective at daily time scales or higher during the GPM era. Finally, a commonly used spatial interpolation technique (kriging), that leverages the spatial correlation of estimation uncertainty, is assessed at climatologic, seasonal, monthly and weekly timescales. It is found that the effectiveness of kriging is sensitive to the

  9. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  10. A southern region conference on technology transfer and extension

    Science.gov (United States)

    Sarah F. Ashton; William G. Hubbard; H. Michael Rauscher

    2009-01-01

    Forest landowners and managers have different education and technology transfer needs and preferences. To be effective it is important to use a multi-faceted science delivery/technology transfer program to reach them. Multi-faceted science delivery programs can provide similar content over a wide range of mechanisms including printed publications, face-to-face...

  11. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  12. The Study of Impacts of Water Transferring From Wet Regions To Dry Regions In Iran

    Science.gov (United States)

    Motiee-Homayoun, Dr.; Ghomashchi, Dr.

    available. In this situation, water transformation from wet areas (with good water resources) to dried and desert regions of the country has been identified as a necessary and reasonable policy to tacklewater shortage. Mediterain climate and mountains in north, west and southwest regions of Iran grant a benefit of high level rate of rainfall, several deep and long rivers, and large capacity of groundwater resources in these areas. Existence of such rivers and water resources, especially a big river of Karoon in southwest, strengthens the goal of constructing hydraulic structures in order to transfer water fro m wet areas to central and eastern areas of the country. This goal has led to planning and implementing of several large and high cost projects. Experts of water affairs, believe that although drinking water supply is one of the most crucial missions of the government, it should also be noted that transformation huge amount of water from an area to another area, with a very long distant, undoubtedly, will cause significant environmental impacts in future. Therefore, decision making and implementing such strategic projects needs a very precise consideration and accurate cost-benefit analyzes. On the one hand, through a socio- economic approach, implementation of such big projects for water transferring requires a great amount of investment and a long period to complete, and benefit peoples. So in many cases multi-purpose and multi- dimensional projects should be considered carefully. On the other hand, water supply for some provinces is vital. In most identified areas, water scarcity is the main cause of urban decline, economic problems and finally loosing population because of emigration. Thus, fresh water should be supplied for these provinces at the earliest possible. This paper is an attempt to identify, define and explain the characteristics and specification of all projects for transferring in different parts of Iran. Generally, advantages and disadvantages of

  13. Dissemination of CERN's Technology Transfer: Added Value from Regional Transfer Agents

    Science.gov (United States)

    Hofer, Franz

    2005-01-01

    Technologies developed at CERN, the European Organization for Nuclear Research, are disseminated via a network of external technology transfer officers. Each of CERN's 20 member states has appointed at least one technology transfer officer to help establish links with CERN. This network has been in place since 2001 and early experiences indicate…

  14. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  15. Modeling chemistry in PWR fuel crud

    International Nuclear Information System (INIS)

    PWR fuel crud arises from deposition of corrosion products in the coolant on clad surfaces in the core. These deposits form a porous layer through which water must pass to provide effective heat transfer from the clad surface. The usual heat transfer mechanism is by wick boiling in which water passing through the porous crud is converted to steam that escapes through steam chimneys in the deposit. This conversion of water into steam within the deposit means that dilute solutions in the bulk coolant become concentrated in the crud and this can lead to precipitation of species such as lithium borates, ZnO and Zn-silicates. Such precipitation processes may lead to problems such as crud induced power shifts (CIPS), formally known as axial off-set anomaly (AOA), or crud induced localised corrosion (CILC) of the clad, that has led to cladding failures. These precipitation processes also hinder heat transfer and can lead to hot spots on the clad surfaces that are potentially damaging. Questions such as what should be the plant limits on Zn, Si, B and Li to prevent such problems, and how should these be controlled during the cycle, are not easy to answer. With several new designs of PWR proposing high power density cores and therefore greater subcooled nucleate boiling, and with existing plants still up-rating their cores, these questions are likely to become more important in the future. It is therefore important to understand the relationship between coolant chemistry (Zn, Si, Li, B levels) and the chemistry within fuel crud. The bulk and crud chemistry are coupled along with the bulk and local heat transfer processes. This coupling of chemistry and heat transfer makes this a particularly difficult problem to investigate theoretically although the authors have previously achieved this using a number of one dimensional heat and mass transfer models. This paper discusses a new approach to this problem using finite element methods to solve the relevant coupled chemistry and

  16. Heat transfer in the post dryout region and on wetting heated surfaces

    International Nuclear Information System (INIS)

    Rassokhin, N.G.; Kabanov, L.P.

    1987-01-01

    A survey is given of the works published in the Soviet Union during 1983 and 1984 on heat transfer in the post dryout region and on wetting heated surfaces. New experimental data, heat transfer models, and computational techniques are analysed. The complexities of the heat transfer process under the above conditions are noted. The differences and common features of the heat transfer processes in the post dryout region and on wetting heated surfaces are indicated as well as the necessity for the development of computational techniques that would consider the two processes simultaneously. (author)

  17. Behavior of heat transfer in pulsating flow in the channel entrance region

    Science.gov (United States)

    Davletshin, I. A.; Paereliy, A. A.

    2017-09-01

    Convective heat transfer in the entrance region of a plane channel has been studied experimentally. Steady and pulsating air flow regimes have been considered. Heat transfer distributions over the channel wall have been obtained. Non-monotonous behavior of heat transfer coefficient has been revealed at some frequencies and amplitudes of forced flow pulsations. It was accompanied by heat transfer enhancement that was almost double for the local heat transfer coefficients. This may result from generation of vortices in the inlet duct in conditions of flow pulsations.

  18. Neutron transfer reactions in the fp-shell region

    International Nuclear Information System (INIS)

    Mahgoub, Mahmoud

    2008-01-01

    Neutron transfer reactions were used to study the stability of the magic number N=28 near 56 Ni. On one hand the one-neutron pickup (d,p) reaction was used for precision spectroscopy of single-particle levels in 55 Fe. On the other hand we investigated the two-neutron transfer mechanism into 56 Ni using the pickup reaction 58 Ni(vectorp,t) 56 Ni. In addition the reliability of inverse kinematics reactions at low energy to study exotic nuclei was tested by the neutron transfer reactions t( 40 Ar,p) 42 Ar and d( 54 Fe,p) 55 Fe using tritium and deuterium targets, respectively, and by comparing the results with those of the normal kinematics reactions. The experimental data, differential cross-section and analyzing powers, are compared to DWBA and coupled channel calculations utilizing the code CHUCK3. By performing the single-neutron stripping reaction (vectord,p) on 54 Fe the 1f 7/2 shell in the ground state configuration was found to be partly broken. The instability of the 1f 7/2 shell and the magic number N=28 was confirmed once by observing a number of levels with J π = 7/2 - at low excitation energies, which should not be populated if 54 Fe has a closed 1f 7/2 shell, and also by comparing our high precision experimental data with a large scale shell model calculation using the ANTOINE code [5]. Calculations including a partly broken 1f 7/2 shell show better agreement with the experiment. The instability of the 1f 7/2 shell was confirmed also by performing the two-neutron pick-up reaction (vectorp,t) on 58 Ni to study 56 Ni, where a considerable improvement in the DWBA calculation was observed after considering 1f 7/2 as a broken shell. To prove the reliability of inverse kinematics transfer reactions at low energies (∝ 2 AMeV), the aforementioned single-neutron transfer reaction (d,p) was repeated using a beam of 54 Fe ions and a deuteron target. From this inverse kinematics experiment we were able to reproduce the absolute cross-section and angular

  19. Historic transfer of forest reproductive material in the Nordic region

    DEFF Research Database (Denmark)

    Myking, Tor; Rusanen, Mari; Steffenrem, Arne

    2016-01-01

    and gene pools of key forest tree species. We find that large imports of non-native FRM occurred from the 19th century onwards, partly due to prior deforestations directly associated with charcoal production for mining, extraction of timber and production of tar and pitch which have historically been......Large-scale transfer of reproductive material is a common phenomenon in forestry and is not only limited to recent history. Here we review the historical transfer of forest reproductive material (FRM) in Fennoscandia, the directions, their drivers, and the reported consequences for adaptation...... important export commodities for Sweden, Norway, and Finland. In Denmark, conversion to agricultural land and the use of forests for livestock feeding was similarly important. During the subsequent reforestation efforts in Denmark, the introduction and use of non-autochthonous FRM of beech, oak and Scots...

  20. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

  1. 75 FR 66702 - Western Electric Coordinating Council; Qualified Transfer Path Unscheduled Flow Relief Regional...

    Science.gov (United States)

    2010-10-29

    ... Commission (Commission) proposes to approve regional Reliability Standard IRO-006-WECC-1 (Qualified Transfer... Notice of Proposed Rulemaking, IRO-006-WECC-1 raises some concerns about which the Commission requests... Standard IRO-006- WECC-1 (Qualified Transfer Path Unscheduled Flow Relief) submitted to the Commission for...

  2. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  3. Overview of PWR chemistry options

    Energy Technology Data Exchange (ETDEWEB)

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  4. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  5. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  6. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  7. 76 FR 69736 - Primus Solutions, Inc., and Arctic Slope Regional Corp.; Transfer of Data

    Science.gov (United States)

    2011-11-09

    ... AGENCY Primus Solutions, Inc., and Arctic Slope Regional Corp.; Transfer of Data AGENCY: Environmental... Rodenticide Act (FIFRA) and the Federal Food, Drug, and Cosmetic Act (FFDCA), including information that may... Primus Solutions, Inc., and its subcontractor, Arctic Slope Regional Corporation (ASRC) in ] accordance...

  8. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  9. Urban and rural patterns in emergent pediatric transfer: a call for regionalization.

    Science.gov (United States)

    Horeczko, Timothy; Marcin, James P; Kahn, Jeremy M; Sapien, Robert E

    2014-01-01

    National groups call for the regionalization of health care, to direct patients with high-risk conditions to designated hospitals with greater capabilities. Currently there is limited information detailing the characteristics and specific needs of acutely ill and injured children who require transfer to another institution, especially in underserved rural communities. To determine the epidemiology of pediatric transfers from urban and rural emergency departments (EDs). We analyzed data in the National Hospital Ambulatory Medical Care Survey from 1995 to 2010. Eligible children were Urban and rural EDs showed similar transfer rates. Children transferred from rural EDs were older and more likely to arrive by emergency medical services than children transferred from urban EDs (12.1 vs 8.2 years of age, P rural EDs were more than twice as likely to be transferred for a psychiatric indication (43.5% vs 19.5%, P urban and rural settings. Rural children have additional obstacles to care, especially in access to emergency mental health services. Programs to study and implement regionalization of care should consider diverse patient populations and target improvement in coordination of care, transfer times, and outcomes. © 2013 National Rural Health Association.

  10. Transcriptional analysis of the leading region in F plasmid DNA transfer.

    Science.gov (United States)

    Cram, D; Ray, A; O'Gorman, L; Skurray, R

    1984-05-01

    Transcriptional activity associated with the leading region (53.8-66.7F) in F DNA transfer has been shown by RNA-DNA hybridization studies to occur on the anterior segment extending from 59.4 to 66.7F. Promoter-probe analysis of cloned leading region segments detected two promoters within the transcribed portion of the leading region. The promoter active across the 64.7F EcoRI site on the transferred F strand was associated with the expression of two polypeptides, 6d and 13.5p, located between 64.7-66.6F. However, no definite role could be ascribed to the second promoter operative through the 66.6F Bg/II site located in close proximity to oriT, the origin of transfer.

  11. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  12. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

  13. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    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

  14. 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.

  15. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  16. Scaling Analysis for PWR Steam Generator

    International Nuclear Information System (INIS)

    Li, Yuquan; Ye, Zishen

    2011-01-01

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly

  17. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    reactor coolant system, combustible gas burning, atmosphere- structure heat transfer, ventilation, containment spray cooling, etc. After the molten core penetrates the reactor bottom head, steam generation, concrete disintegration and noncondensable gas generation are calculated in the reactor cavity or the pedestal. 2 - Method of solution: Each of the THALES member codes first establishes the steady state conditions after reading input data. Then iterative time-dependent calculation is continued, taking account of various phenomena and events and their interactions which will occur in the course of a postulated severe accident. The transient calculations are iterated by the physical times specified by input. Generally the RCS thermal hydraulic analysis with the THALES-PM or THALES-BM code is first carried out and its results are transferred to the following containment analysis with the THALES-CV code. Then both results are transferred to a code for analyzing fission product release and transport behavior. Automatic data transfer is possible in the case the JAERI's ART code is used for fission product behavior analysis. In overall thermal hydraulic analysis, a new method is adopted aiming at sufficiently accurate estimation of mixture levels in the reactor coolant system and the containment in a reasonable computer time. The heat transfer calculation in the core is carried out based on the backward method. 3 - Restrictions on the complexity of the problem: Restrictions relating to storage allocation are: (1) Maximum number of radial regions in the core : 10; (2) Maximum number of axial increments in the fuel rods : 50; (3) Maximum number of loops in the PWR primary system : 4; (4) Maximum number of volumes in the PWR primary system : 11; (5) Number of BWR recirculation loops: 2 (fixed); (6) Number of volumes in the BWR reactor coolant system : 7 (fixed); (7) Maximum number of compartments in the containment : 10. There is another restriction, which relates to time step

  18. 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)

  19. Long-range electron transfer in engineered azurins exhibits marcus inverted region behavior

    DEFF Research Database (Denmark)

    Farver, Ole; Hosseinzadeh, Parisa; Marshall, Nicholas M.

    2015-01-01

    The Marcus theory of electron transfer (ET) predicts that while the ET rate constants increase with rising driving force until it equals a reaction’s reorganization energy, at higher driving force the ET rate decreases, having reached the Marcus inverted region. While experimental evidence...

  20. Examining the effects of parameter regionalization schemes on parameter transferability on large basin sampling

    Science.gov (United States)

    Rakovec, Oldrich; Mizukami, Naoki; Newman, Andrew; Thober, Stephan; Kumar, Rohini; Wood, Andrew; Clark, Martyn P.; Samaniego, Luis

    2017-04-01

    Assessing model complexity and performing "seamless" continental-domain model simulations (e.g., model parameters yielding good performance across entire domain) is a challenging topic in contemporary hydrology. This study presents a large-sample hydrologic modeling effort to examine the effects of parameter regionalization schemes. Two hydrological models (mHM, VIC) are set up for 500 small to medium-sized unimpaired basins over the contiguous United States for two spatial scales: lumped and 12km grid. For parameter regionalization, we use the well-established Multiscale Parameter Regionalization (MPR) technique for both models, with the specific goal of assessing the transferability of model parameters across different spatial scales (lumped basin scale to distributed), time periods (from calibration to validation period), and locations. In terms of the scale transferability, evaluation of global model parameters at finer scale based on calibration at coarse scale improves the KGE performance (mainly due to the variance related term). Loss in model performance in temporal transferability is independent from model complexity (i.e., lumped vs. distributed). Finally, we show that although the parameter regionalization is crucial for parameter transferability to un-gauged locations, there still remains room for improvement especially for the mean and variability in streamflow. We present possible strategies to resolve this issue, including (1) assessing the importance of more detailed information on the soil data (STATSGO vs. SoilGrids), and (2) applying more advanced selection criteria for training MPR global parameters.

  1. Control of a hybrid HVDC link to increase inter-regional power transfer

    DEFF Research Database (Denmark)

    Kotb, Omar; Ghandhari, Mehrdad; Eriksson, Robert

    2016-01-01

    This paper examines the application of a hybrid HVDC link in a two area power system with the purpose of increasing the inter-regional power transfer. A hybrid HVDC system combines both LCCs and VSCs, and hence it is capable of combining the benefits of both converter technologies, such as reduced...

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

    International Nuclear Information System (INIS)

    Iguchi, Tadashi

    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

  3. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  4. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  5. Regional Cooperation Agreement for Asia and the Pacific (RCA). A mechanism for nuclear technology transfer

    International Nuclear Information System (INIS)

    Bin Muslim, N.

    1993-01-01

    The paper presents the regional cooperation programs of the IAEA which have as purpose to promote the applications of peaceful uses of atomic energy and to transfer technology to the developing countries. The paper focusses on the (RCA) program for Asia and the Pacific, it is considered the most important mechanism for genuine technology transfer. The annex no 1 lists the full text of the Regional Cooperative Agreement for Research, Development and Training Related to Nuclear Science and Technology, 1987 (13 articles). The annex no.3 lists also the full text of the African Regional Cooperative Agreement for Research, Development and training Related to Nuclear Science and Technology (14 articles). 11 refs., 17 tabs

  6. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  7. Lateral gene transfer of streptococcal ICE element RD2 (region of difference 2 encoding secreted proteins

    Directory of Open Access Journals (Sweden)

    Mereghetti Laurent

    2011-04-01

    Full Text Available Abstract Background The genome of serotype M28 group A Streptococcus (GAS strain MGAS6180 contains a novel genetic element named Region of Difference 2 (RD2 that encodes seven putative secreted extracellular proteins. RD2 is present in all serotype M28 strains and strains of several other GAS serotypes associated with female urogenital infections. We show here that the GAS RD2 element is present in strain MGAS6180 both as an integrative chromosomal form and a circular extrachromosomal element. RD2-like regions were identified in publicly available genome sequences of strains representing three of the five major group B streptococcal serotypes causing human disease. Ten RD2-encoded proteins have significant similarity to proteins involved in conjugative transfer of Streptococcus thermophilus integrative chromosomal elements (ICEs. Results We transferred RD2 from GAS strain MGAS6180 (serotype M28 to serotype M1 and M4 GAS strains by filter mating. The copy number of the RD2 element was rapidly and significantly increased following treatment of strain MGAS6180 with mitomycin C, a DNA damaging agent. Using a PCR-based method, we also identified RD2-like regions in multiple group C and G strains of Streptococcus dysgalactiae subsp.equisimilis cultured from invasive human infections. Conclusions Taken together, the data indicate that the RD2 element has disseminated by lateral gene transfer to genetically diverse strains of human-pathogenic streptococci.

  8. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  9. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  10. Transfer factors for the „soil-cereals” system in the region of Pcinja, Serbia

    Directory of Open Access Journals (Sweden)

    Marković Jelena S.

    2016-01-01

    Full Text Available The aim of the paper was to estimate the values of transfer factors for natural radionuclides (40K, 226Ra, 232Th, 235U, and 238U and 137Cs from soil to plants (cereals: wheat, corn and barley as important parameters for the agricultures in the selection of the location and the sort of cereals to be planted on. The results presented in this paper refer to the „soil-cereals” system in the region of Pcinja, Serbia. Total of 9 samples of soil and 7 samples of cereals were measured in the Department of Radiation and Environmental Protection, Vinca Institute of Nuclear Sciences, using three high-purity germanium detectors for gamma spectrometry measurements. In all the samples, transfer factors for 226Ra are significantly lower than for 40K, but they are all in good agreement with the literature data. On the three investigated locations, the calculated values of transfer factors for 40K were in the range of 0.144 to 0.392, while in the case of 226Ra, the transfer factors ranged from 0.008 to 0.074. Only one value (0.051 was obtained for transfer factor of 232Th. Specific activities of 137Cs, as well as uranium isotopes, in all the investigated cereal samples, were below minimal detectable activity concentrations. Also, the absorbed dose rate and the annual absorbed dose from the natural radionuclides in the soil, were calculated. The absorbed dose rate ranged from 49-86 nSv/h, while the annual absorbed dose ranged from 0.061-0.105 mSv. The measurements presented in this manuscript are the first to be conducted in the region of Pcinja, thus providing the results that can be used as a baseline for future measurements and monitoring.

  11. Beta and gamma dose calculations for PWR and BWR containments

    International Nuclear Information System (INIS)

    King, D.B.

    1989-07-01

    Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 x 10 8 rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 x 10 8 rad equipment qualification test region. 8 refs., 23 figs., 12 tabs

  12. York Region integrated solid waste processing and transfer facility : cash flow analysis of alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Balfour, B. [Gartner Lee Ltd., Markham, ON (Canada)

    2000-07-01

    Cash flow modeling exercises of the two tier waste management system in York Region was presented to encourage decision-makers to think in terms of total system costs and how to achieve a desired diversion rate. The York Region consists of 9 municipalities which collect their own waste. The region is responsible for its treatment and disposal. This paper have shown that modeling of realistic options gives decision-makers the opportunity to see the financial impact of different types of waste management systems and provides them with the potential to view the critical parameters in the cost of waste management. This paper demonstrated through modelling the impacts of various alternatives such as collection, processing, transfer and disposal, and identified achievable diversions.

  13. Stabilizing Parametric Region of Multiloop PID Controllers for Multivariable Systems Based on Equivalent Transfer Function

    Directory of Open Access Journals (Sweden)

    Xiaoli Luan

    2016-01-01

    Full Text Available The aim of this paper is to determine the stabilizing PID parametric region for multivariable systems. Firstly, a general equivalent transfer function parameterization method is proposed to construct the multiloop equivalent process for multivariable systems. Then, based on the equivalent single loops, a model-based method is presented to derive the stabilizing PID parametric region by using the generalized Hermite-Biehler theorem. By sweeping over the entire ranges of feasible proportional gains and determining the stabilizing regions in the space of integral and derivative gains, the complete set of stabilizing PID controllers can be determined. The robustness of the design procedure against the approximation in getting the SISO plants is analyzed. Finally, simulation of a practical model is carried out to illustrate the effectiveness of the proposed technique.

  14. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  15. PWR Analysis with the Advanced System: DELFOS

    International Nuclear Information System (INIS)

    Cabellos, O.; Aragones, J.M.; Ahnert, C.

    1998-01-01

    The development of new PWR codes is necessary due to the heterogeneity of fuel assemblies, the complexity of load patterns and the required operation conditions. Code revisions have been previously referred. Although modern advanced nodal core models have been well established, some reports in the Annual Conference of the A.N.S. in 1995 indicated that the accuracy of cross section models have received less attention. Due to the new performance and taking into account the importance of the nodal cross-sections approximations, the group of researchers in the Instituto de Fusion Nuclear (UPM)have developed new models (code systems DELFOS) for advanced analysis of PWR cores. The system has been tested in the Asco II NPP, cycle 1 to 11 (nominal operation and startup physics tests) comparing with measurements in the last cycle. In conclusion we have validated this methodology for its general application to PWR reactors. (Author)

  16. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  17. Turbulent convective heat and mass transfer in the developing region of elliptical ducts

    International Nuclear Information System (INIS)

    Vinagre, H.T.M.; Mendes, P.R.S.

    1990-01-01

    Mass transfer experiments were performed to determine local heat and mass transfer coefficients for the turbulent flow in a duct with elliptical cross section. The naphthalene sublimation technique was employed to obtain the experimental results. Both entrance-region and fully-developed results were obtained. The Reynolds number was varied in the overall range of 7000-60,000, whereas values of 0,12, 0,25 and 0,5 for the aspect ratio were investigated. The fully developed transport coefficients obtained were compared with the ones available in the open literature for parallel plates and circular tubes, and it was found that the coefficients are quite insensitive to aspect ratio variations. (author)

  18. Inclusive electron scattering from nuclei in the quasielastic region at large momentum transfer

    Energy Technology Data Exchange (ETDEWEB)

    Fomin, Nadia [California Inst. of Technology (CalTech), Pasadena, CA (United States)

    2008-12-01

    Experiment E02-019, performed in Hall C at the Thomas Jefferson National Accelerator Facility (TJNAF), was a measurement of inclusive electron cross sections for several nuclei (2H,3He, 4He, 9Be,12C, 63Cu, and 197Au) in the quasielastic region at high momentum transfer. In the region of low energy transfer, the cross sections were analyzed in terms of the reduced response, F(y), by examining its y-scaling behavior. The data were also examined in terms of the nuclear structure function νWA 2 and its behavior in x and the Nachtmann variable ξ. The data show approximate scaling of νWA 2 in ξ for all targets at all kinematics, unlike scaling in x, which is confined to the DIS regime. However, y-scaling observations are limited to the kinematic region dominated by the quasielastic response (y <0), where some scaling violations arising from FSIs are observed.

  19. Discrete transfer method with the concept of blocked-off region for irregular geometries

    Energy Technology Data Exchange (ETDEWEB)

    Talukdar, Prabal [Institute of Fluid Mechanics (LSTM), University of Erlangen-Nuremberg, Cauerstrasse 4, D-91058 Erlangen (Germany)]. E-mail: prabal_iitg@yahoo.com

    2006-03-15

    The discrete transfer method (DTM) is applied to irregular geometries with a concept of blocked-off region previously applied in the problems of computational fluid dynamics. This gives a new alternative to the DTM for its implementation to irregular structures. The Cartesian coordinate-based ray-tracing algorithm can be applied to the geometries with inclined or curved boundaries. Some test problems are considered and results are validated with the available results in the literature. Both radiative and non-radiative equilibrium situations are considered. The medium is assumed to be both participating and non-participating. Results are found to be accurate for all kinds of situations.

  20. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  1. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    International Nuclear Information System (INIS)

    Akiyama, Mamoru; Inoue, Akira; Miyazaki, Keiji; Abeta, Sadaaki; Hori, Keiichi; 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

  2. Transfer coefficient of 137Cs from feed to cow milk in tropical region Kaiga (India)

    International Nuclear Information System (INIS)

    Joshi, R. M.; James, J. P.; Dileep, B. N.; Mulla, R. M.; Reji, T. K.; Ravi, P. M.; Hegde, A. G.; Sarkar, P. K.

    2012-01-01

    In the transport model for the prediction of the concentration of 137 Cs in milk, the transfer coefficient from feed to milk, F m , is an important parameter. Site-specific transfer coefficient from feed to cow's milk, for 137 Cs in the Kaiga environment, a nuclear power station site in India, determined over a period of 10 y is presented in this paper. The value is determined from 137 Cs concentration in milk and grass samples of the Kaiga region and the result ranged from 6.43 E-03 to 1.09 E-02 d l -1 with a geometric mean value of 8.0 E-03 d l -1 . The result is compared with that for 40 K, determined concurrently at the same region and ranged from 3.06 E-03 to 3.48 E-03 d l -1 with a geometric mean value of 3.26 E-03 d l -1 . This parameter is quite useful in decision-making for implementing countermeasures during a large area contamination with 137 Cs in tropical areas like Kaiga. (authors)

  3. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  4. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  5. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  6. 3D graphics simulation of the PWR

    International Nuclear Information System (INIS)

    Lei Gongchao; Ma Baiyong

    1999-01-01

    Using the functions of the software 'I-DEAS Master Series 5', such as the mode of design, drafting, simulation, test, geometry and so on, the task of stereo graphics simulating the PWR is done. Reliability of designed data is checked

  7. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  8. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  9. Manufacturing technologies of PWR pressure vessels

    International Nuclear Information System (INIS)

    Qin Xubin

    1991-01-01

    Pressure vessels belong to the main component of PWR plants. Starting with describing the manufacture of pressure vessel components and their assembly, the manufacturing technologies of pressure vessels are briefly presented with regards to welding, heat treatment, inspections and testing. In addition, quality assurance during the manufacture is presented with emphasis

  10. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  11. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1978-06-01

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

  12. Parameter transferability within homogeneous regions and comparisons with predictions from a priori parameters in the eastern United States

    Science.gov (United States)

    Chouaib, Wafa; Alila, Younes; Caldwell, Peter V.

    2018-05-01

    The need for predictions of flow time-series persists at ungauged catchments, motivating the research goals of our study. By means of the Sacramento model, this paper explores the use of parameter transfer within homogeneous regions of similar climate and flow characteristics and makes comparisons with predictions from a priori parameters. We assessed the performance using the Nash-Sutcliffe (NS), bias, mean monthly hydrograph and flow duration curve (FDC). The study was conducted on a large dataset of 73 catchments within the eastern US. Two approaches to the parameter transferability were developed and evaluated; (i) the within homogeneous region parameter transfer using one donor catchment specific to each region, (ii) the parameter transfer disregarding the geographical limits of homogeneous regions, where one donor catchment was common to all regions. Comparisons between both parameter transfers enabled to assess the gain in performance from the parameter regionalization and its respective constraints and limitations. The parameter transfer within homogeneous regions outperformed the a priori parameters and led to a decrease in bias and increase in efficiency reaching a median NS of 0.77 and a NS of 0.85 at individual catchments. The use of FDC revealed the effect of bias on the inaccuracy of prediction from parameter transfer. In one specific region, of mountainous and forested catchments, the prediction accuracy of the parameter transfer was less satisfactory and equivalent to a priori parameters. In this region, the parameter transfer from the outsider catchment provided the best performance; less-biased with smaller uncertainty in medium flow percentiles (40%-60%). The large disparity of energy conditions explained the lack of performance from parameter transfer in this region. Besides, the subsurface stormflow is predominant and there is a likelihood of lateral preferential flow, which according to its specific properties further explained the reduced

  13. Evidences of dynamic stress transfer at Mt. Etna volcano by regional and teleseismic earthquakes

    Science.gov (United States)

    Cannata, Andrea; di Grazia, Giuseppe; Montalto, Placido; Aliotta, Marco; Patanè, Domenico

    2010-05-01

    Influences of distant earthquakes on volcanic systems by dynamic stress transfer are well documented. We analysed seismic signals and volcanic activity at Mt. Etna during time periods characterised by strong regional and teleseismic earthquakes. Two periods, January 2006 and May 2008 showed variations clearly time-related to distant earthquakes. In the first period, characterised by mild volcano activity, the effect of the dynamic stress transfer, caused by a Greek earthquake (M=6.8), was duplex: i) banded tremor activity, whose source is generally related to the existence of a shallow hydrothermal system, changed its features and almost disappeared; ii) a swarm of volcano-tectonic earthquakes with focal depth of 10-15 km b.s.l. took place. The changes of the banded tremor were related to variations in parameters, such as heat and steam flows, permeability and porosity of the rocks, likely caused by weak dynamic stresses. On the other hand, VT earthquake swarm probably developed as a secondary process, promoted by the dynamically triggered activation of magmatic fluids. The second period, May 2008, showed an intense explosive activity. During this time interval the dynamic stress transfer, related to the arrival of the seismic waves of the Sichuan earthquake (M=7.9), affected the features of the seismo-volcanic signals and triggered an eruption on the following day. In particular, we observed a gradual decrease of volcanic tremor amplitude, just after the arrival of the teleseismic waves, related to tremor source shift, and an increase of both occurrence rate and energy of long period events. In this case, we suggest that dynamic stress transfer caused buildup of pressure in magma bodies, that was highlighted by the increase of LP activity. On the following day such increasing overpressure led to an eruption. In conclusion, phenomena of dynamic stress transfer have been recognized at Mt. Etna able to modify the state of the volcano. However, uncertainties still

  14. Effects of Micro-fin Structure on Spray Cooling Heat Transfer in Forced Convection and Nucleate Boiling Region

    International Nuclear Information System (INIS)

    Kim, Yeung Chan

    2010-01-01

    In the present study, spray cooling heat transfer was experimentally investigated for the case in which water is sprayed onto the surfaces of micro-fins in forced convection and nucleate boiling regions. The experimental results show that an increase in the droplet flow rate improves heat transfer due to forced convection and nucleate boiling in the both case of smooth surface and surfaces of micro-fins. However, the effect of subcooling for fixed droplet flow rate is very weak. Micro-fins surfaces enhance the spray cooling heat transfer significantly. In the dilute spray region, the micro-fin structure has a significant effect on the spray cooling heat transfer. However, this effect is weak in the dense spray region. A previously determined correlation between the Nusselt number and Reynolds number shows good agreement with the present experimental data for a smooth surface

  15. Heat transfer--Orlando (Symposium), 1980

    International Nuclear Information System (INIS)

    Stein, R.P.

    1980-01-01

    This conference proceedings contains 36 papers of which 3 appear as abstracts. 23 papers are indexed separately. Topics covered include: thermodynamics of PWR and LMFBR Steam Generators; two-phase flow in parallel channels; geothermal heat transfer; natural circulation in complex geometries; heat transfer in non-Newtonian systems; and process heat transfer

  16. Development of a Fast and Accurate PCRTM Radiative Transfer Model in the Solar Spectral Region

    Science.gov (United States)

    Liu, Xu; Yang, Qiguang; Li, Hui; Jin, Zhonghai; Wu, Wan; Kizer, Susan; Zhou, Daniel K.; Yang, Ping

    2016-01-01

    A fast and accurate principal component-based radiative transfer model in the solar spectral region (PCRTMSOLAR) has been developed. The algorithm is capable of simulating reflected solar spectra in both clear sky and cloudy atmospheric conditions. Multiple scattering of the solar beam by the multilayer clouds and aerosols are calculated using a discrete ordinate radiative transfer scheme. The PCRTM-SOLAR model can be trained to simulate top-of-atmosphere radiance or reflectance spectra with spectral resolution ranging from 1 cm(exp -1) resolution to a few nanometers. Broadband radiances or reflectance can also be calculated if desired. The current version of the PCRTM-SOLAR covers a spectral range from 300 to 2500 nm. The model is valid for solar zenith angles ranging from 0 to 80 deg, the instrument view zenith angles ranging from 0 to 70 deg, and the relative azimuthal angles ranging from 0 to 360 deg. Depending on the number of spectral channels, the speed of the current version of PCRTM-SOLAR is a few hundred to over one thousand times faster than the medium speed correlated-k option MODTRAN5. The absolute RMS error in channel radiance is smaller than 10(exp -3) mW/cm)exp 2)/sr/cm(exp -1) and the relative error is typically less than 0.2%.

  17. Heat transfer in the entrance region of symmetric and asymmetric finite circular rod arrays

    International Nuclear Information System (INIS)

    Sengupta, S.; Narasimhan, R.

    1987-01-01

    Heat transfer in the combined entrance region of symmetric and asymmetric finite circular rod bundles is solved using the boundary fitted coordinate system. It is found that in symmetric bundles the fully developed and the developing local bundle Nusselt number increases with the peripheral rod radius to a maximum after which it decreases. Three types of eccentric bundles are studied. Large displacement in rod leads to decrease in the fully developed and the developing local bundle Nusselt number. However, small eccentricities in bundle with peripheral rod radius smaller than the one at which the maximum bundle Nusselt number occurs, lead to slight increases in the bundle Nusselt number. Even small eccentricities of the rods affect the entire temperature field of the bundle. This is unlike the velocity field which is affected only in the neighborhood of the displaced rod

  18. China’s regional CH4 emissions: Characteristics, interregional transfer and mitigation policies

    International Nuclear Information System (INIS)

    Zhang, Bo; Yang, T.R.; Chen, B.; Sun, X.D.

    2016-01-01

    Highlights: • China’s CH 4 emissions have significant contributions to global climate change. • The total CH 4 emissions in 2010 amount to 44.3 Tg, half from energy activities. • Half of the national total direct emissions are embodied in interregional trade. • 2/3 of the embodied emission transfers via domestic trade are energy-related. • A national comprehensive action plan to reduce CH 4 emissions should be designed. - Abstract: Methane (CH 4 ), the second largest greenhouse gas emitted in China, hasn’t been given enough attention in the country’s policies and actions for addressing climate change. This paper aims to perform a bottom-up estimation and multi-regional input–output analysis for China’s anthropogenic CH 4 emissions from both production-based and consumption-based insights. As the world’s largest CH 4 emitter, China’s total anthropogenic CH 4 emissions in 2010 are estimated at 44.3 Tg and correspond to 1507.9 Mt CO 2 -eq by the lower global warming potential factor of 34. Energy activities as the largest contributor hold about half of the national total emissions, mainly from coal mining. Inherent economic driving factors covering consumption, investment and international exports play an important role in determining regional CH 4 emission inventories. Interregional transfers of embodied emissions via domestic trade are equivalent to half of the national total emissions from domestic production, of which two thirds are energy-related embodied emissions. Most central and western regions as net interregional CH 4 exporters such as Shanxi and Inner Mongolia have higher direct emissions, while the eastern coastal regions as net interregional importers such as Guangdong and Jiangsu always have larger embodied emissions. Since China’s CH 4 emissions have significant contributions to global climate change, a national comprehensive action plan to reduce CH 4 emissions should be designed by considering supply-side and demand

  19. Progress of PWR reactor fuels: OSIRIS equipments

    International Nuclear Information System (INIS)

    Colomez, G.; Farny, G.; Vidal, H.

    1981-09-01

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering [fr

  20. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  1. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  2. Horizontal Drop of 21- PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    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

  3. On the comparability of knowledge transfer activities - a case study at the German Baltic Sea Coast focusing regional climate services

    Science.gov (United States)

    Meinke, Insa

    2017-06-01

    In this article the comparability of knowledge transfer activities is discussed by accounting for external impacts. It is shown that factors which are neither part of the knowledge transfer activity nor part of the participating institution may have significant impact on the potential usefulness of knowledge transfer activities. Differences in the potential usefulness are leading to different initial conditions of the knowledge transfer activities. This needs to be taken into account when comparing different knowledge transfer activities, e.g., in program evaluations. This study is focusing on regional climate services at the German Baltic Sea coast. It is based on two surveys and experiences with two identical web tools applied on two regions with different spatial coverage. The results show that comparability among science based knowledge transfer activities is strongly limited through several external impacts. The potential usefulness and thus the initial condition of a particular knowledge transfer activity strongly depends on (1) the perceived priority of the focused topic, (2) the used information channels, (3) the conformity between the research agenda of service providing institutions and information demands in the public, as well as (4) on the spatial coverage of a service. It is suggested to account for the described external impacts for evaluations of knowledge transfer activities. The results show that the comparability of knowledge transfer activities is limited and challenge the adequacy of quantitative measures in this context. Moreover, as shown in this case study, in particular regional climate services should be individually evaluated on a long term perspective, by potential user groups and/or by its real users. It is further suggested that evaluation criteria should be co-developed with these stakeholder groups.

  4. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  5. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    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

  6. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  7. Nuclear event time histories and computed site transfer functions for locations in the Los Angeles region

    Science.gov (United States)

    Rogers, A.M.; Covington, P.A.; Park, R.B.; Borcherdt, R.D.; Perkins, D.M.

    1980-01-01

    This report presents a collection of Nevada Test Site (NTS) nuclear explosion recordings obtained at sites in the greater Los Angeles, Calif., region. The report includes ground velocity time histories, as well as, derived site transfer functions. These data have been collected as part of a study to evaluate the validity of using low-level ground motions to predict the frequency-dependent response of a site during an earthquake. For this study 19 nuclear events were recorded at 98 separate locations. Some of these sites have recorded more than one of the nuclear explosions, and, consequently, there are a total of 159, three-component station records. The location of all the recording sites are shown in figures 1–5, the station coordinates and abbreviations are given in table 1. The station addresses are listed in table 2, and the nuclear explosions that were recorded are listed in table 3. The recording sites were chosen on the basis of three criteria: (1) that the underlying geological conditions were representative of conditions over significant areas of the region, (2) that the site was the location of a strong-motion recording of the 1971 San Fernando earthquake, or (3) that more complete geographical coverage was required in that location.

  8. RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

    International Nuclear Information System (INIS)

    Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

    1984-01-01

    Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

  9. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  10. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  11. 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

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

    International Nuclear Information System (INIS)

    Chung, H.M.

    2006-01-01

    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 rates, and

  13. In situ corrosion monitoring of steam generators. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kendig, M W; Isaacs, H S

    1978-06-01

    An ac electrochemical technique which meets the basic requirements for an in situ localized corrosion monitor within the secondary coolant of PWR steam generators has been investigated. The technique uses two electrodes to measure the electrochemical impedance of a surface in an occluded region with high heat flux. The impedance is related to the kinetics of corrosion. Marked decreases indicate the onset of a high corrosion rate. Experiments have demonstrated the ability of the technique to determine the onset of corrosion under conditions of high solution resistance and solution agitation due to local boiling. Experiments have shown the technique operates similarly in pressurized 300/sup 0/C water, 1,400 ppM in Na/sub 2/SO/sub 4/.

  14. 76 FR 16691 - Western Electric Coordinating Council Qualified Transfer Path Unscheduled Flow Relief Regional...

    Science.gov (United States)

    2011-03-25

    ... of the Western Electricity Coordinating Council (WECC) IRO-006-WECC-1 (Qualified Transfer Path... transmission overloads due to unscheduled flow on a transfer path designated by WECC as being qualified for... the Western Electricity Coordinating Council (WECC) IRO-006-WECC-1 (Qualified Transfer Path...

  15. A new model for simulation of pressurizers in PWR power plants

    International Nuclear Information System (INIS)

    Madeira, A.A.

    1981-02-01

    The pressurizer of a PWR type reactor was simulated as a thermodynamical system made up of three regions with movable boundaries. The mechanisms of normal condensation, condensation induced by spray, flashing and heat exchange across the water - steam interface, were studied. Various tests have been carried out and satisfactory results were obtained when compared with those from other models and also with some available experimental data. (E.G.) [pt

  16. The historical place of the PWR in energy supply

    International Nuclear Information System (INIS)

    Rippon, S.

    1982-01-01

    The development of nuclear power and evolutionary changes in PWR technology including plant standardisation are discussed. The proposed Sizewell B nuclear power station would benefit from three sources of standardisation: the architect engineering advice of Bechtel; the licensing package of Westinghouse; and the joint design of the SNUPPS group. Safety issues and PWR performance are also discussed. (U.K.)

  17. Proton-transfer reaction mass spectrometry (PTR-MS) for the authentication of regionally unique South African lamb

    NARCIS (Netherlands)

    Erasmus, Sara W.; Muller, Magdalena; Alewijn, Martin; Koot, Alex H.; Ruth, van Saskia M.; Hoffman, Louwrens C.

    2017-01-01

    The volatile fingerprints of South African lamb meat and fat were measured by proton-transfer mass spectrometry (PTR-MS) to evaluate it as an authentication tool. Meat and fat of the Longissimus lumborum (LL) of lambs from six different regions were assessed. Analysis showed that the volatile

  18. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  19. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  20. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  1. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  2. Fuel cycle cost projections. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Clark, L.L.; Chockie, A.D.

    1979-12-01

    This report estimates current and future costs associated with the light water reactor nuclear fuel cycle for both once-through and thermal recycle cases. Using a range of future nuclear power generating scenarios, process flows are developed for each segment of the nuclear fuel cycle. Capital and operating costs are estimated and are combined with the process flows to generate unit cost projections for each fuel cycle segment. The unit costs and process flows are combined in the NUCOST program to estimate fuel cycle power costs through the year 2020. The unit costs are also used to estimate the fuel costs of an individual model PWR and BWR.

  3. Dust Emission at 8 and 24 μ m as Diagnostics of H ii Region Radiative Transfer

    Energy Technology Data Exchange (ETDEWEB)

    Oey, M. S.; López-Hernández, J.; Kellar, J. A. [Department of Astronomy, University of Michigan, 311 West Hall, 1085 South University Avenue, Ann Arbor, MI, 48109-1107 (United States); Pellegrini, E. W. [Institut für Theoretische Astrophysik, Albert-Überle-Str. 2, D-69120 Heidelberg (Germany); Gordon, K. D.; Meixner, M.; Roman-Duval, J. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States); Jameson, K. E. [Astronomy Department and Laboratory for Millimeter-wave Astronomy, University of Maryland, College Park, MD 20742 (United States); Li, A. [Department of Physics and Astronomy, University of Missouri, Columbia, MO 65211 (United States); Madden, S. C. [Laboratoire AIM, CEA, Université Paris VII, IRFU/Service d’Astrophysique, Bat. 709, F-91191 Gif-sur-Yvette (France); Bot, C. [Observatoire Astronomique de Strasbourg, Université de Strasbourg, CNRS, UMR 7550, 11 Rue de l’Université, F-67000 Strasbourg (France); Rubio, M. [Departamento de Astronomía, Universidad de Chile, Casilla 36-D, Santiago (Chile); Tielens, A. G. G. M. [Leiden Observatory, Leiden University, P.O. Box 9513, NL-2300RA Leiden (Netherlands)

    2017-07-20

    We use the Spitzer Surveying the Agents of Galaxy Evolution (SAGE) survey of the Magellanic Clouds to evaluate the relationship between the 8 μ m polycyclic aromatic hydrocarbon (PAH) emission, 24 μ m hot dust emission, and H ii region radiative transfer. We confirm that in the higher-metallicity Large Magellanic Cloud, PAH destruction is sensitive to optically thin conditions in the nebular Lyman continuum: objects identified as optically thin candidates based on nebular ionization structure show six times lower median 8 μ m surface brightness (0.18 mJy arcsec{sup −2}) than their optically thick counterparts (1.2 mJy arcsec{sup −2}). The 24 μ m surface brightness also shows a factor of three offset between the two classes of objects (0.13 versus 0.44 mJy arcsec{sup −2}, respectively), which is driven by the association between the very small dust grains and higher density gas found at higher nebular optical depths. In contrast, PAH and dust formation in the low-metallicity Small Magellanic Cloud is strongly inhibited such that we find no variation in either 8 μ m or 24 μ m emission between our optically thick and thin samples. This is attributable to extremely low PAH and dust production together with high, corrosive UV photon fluxes in this low-metallicity environment. The dust mass surface densities and gas-to-dust ratios determined from dust maps using Herschel HERITAGE survey data support this interpretation.

  4. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  5. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  6. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    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)

  7. Sensitivity analysis on hot channel of PWR type reactors using matricial formalism

    International Nuclear Information System (INIS)

    Maciel, Edisson Savio G.; Andrade Lima, Fernando Roberto de; Lira, Carlos Alberto B.O.

    1995-01-01

    The matricial formalism of the perturbation theory is applied in a simplified model to study the hot channel of PWR reactors. Mass, linear momentum and energy conservation equations and appropriated heat transfer and fluid mechanics correlations describe the discretized system. After calculating system's thermalhydraulic properties, the matricial formalism is applied and the sensitivity coefficients are determined for each case of interest. Comparisons between perturbative method and direct results of the model have shown good agreement which demonstrates that the matricial formalism is an important tool for discretized system analysis. (author). 6 refs, 2 tabs

  8. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  9. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Mann, C.A.; Hindle, E.D.; Parsons, P.D.

    1982-04-01

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  10. PCARRD's strategies for technology transfer: The agriculture and resources regional technology information system and the regional applied communication program

    International Nuclear Information System (INIS)

    Stuart, T.H.; Mamon, C.R.

    1990-05-01

    This paper describes the Agriculture and Resources Regional Technology Information System (ARRTIS) and the Regional Applied Communication Outreach Program (RAC) of PCARRD. The ARRTIS and the RACO are the strategies in communicating scientific and technology-based information. The ARRTIS is an information system that provides an information base on the status of technologies at various levels of maturity (generation, adaptation, verification, piloting, dissemination and utilization) and offers technology alternatives based on environmental requirements, costs and returns analysis or feasibility of the technologies. This information base provides the repository of technology information from which the Applied Communication Program draws its information for packaging into various formats, using various strategies/media to cater to various users in the regions most especially the farmers. Meanwhile, as PCARRD executes its mission of developing the national research system, it incorporates a development support communication program through the RACO. The RACO is essentially a working component of a regional research center/consortium in each region coordinated by the Applied Communication Division of PCARRD. It aims at reaching farmers and their families, extensionists, administrators, policy makers and entrepreneurs with research information and technology which use a variety of appropriate communication channels, modern communication technology and strategies so that they may actively participate in research diffusion and utilization. (author). 7 refs

  11. The missing dimension of knowledge transfer from subsidiaries to headquarters: The case of Oil and Gas companies in CEE region

    Directory of Open Access Journals (Sweden)

    Emil Velinov

    2016-12-01

    Full Text Available The paper identifies knowledge management determinants of knowledge transfer from subsidiaries to headquarters in the top Oil & Gas companies in Central and Eastern Europe as their level of innovations, internationalization and economic importance are emerging. The paper sheds a light not only on the process of knowledge transfer parent-subsidiary but via versa as it is critical in the 21st century for better adapting to specific business needs in certain geographical regions. Thus, this reversed knowledge from subsidiaries to headquarters is critical for the given business sector where the level of innovation and amount of R&D investments are enormous. The study argues that the reversed process of knowledge transfers from subsidiary to parent company is positively related to company performance and business diversification. Nowadays the knowledge formed in the subsidiaries of Multinational Corporations (MNCs is transferred to headquarters by investing in R&D centres, building new exploration and testing sites abroad. In the reversed knowledge transfer process we can identify main challenges, which are very critical to analyse and determine the exact process.

  12. Radiation heat transfer calculations for the uranium fuel-containment region of the nuclear light bulb engine.

    Science.gov (United States)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1971-01-01

    Calculation results are reviewed of the radiant heat transfer characteristics in the fuel and buffer gas regions of a nuclear light bulb engine based on the transfer of energy by thermal radiation from gaseous uranium fuel in a neon vortex, through an internally cooled transparent wall, to seeded hydrogen propellant. The results indicate that the fraction of UV energy incident on the transparent walls increases with increasing power level. For the reference engine power level of 4600 megw, it is necessary to employ space radiators to reject the UV radiated energy absorbed by the transparent walls. This UV energy can be blocked by employing nitric oxide and oxygen seed gases in the fuel and buffer gas regions. However, this results in increased UV absorption in the buffer gas which also requires space radiators to reject the heat load.

  13. Determination of welding parameters for execution of weld overlayer on PWR nuclear reactor nozzles

    International Nuclear Information System (INIS)

    Ribeiro, Gabriela M.; Lima, Luciana I.; Quinan, Marco A.; Schvartzman, Monica M.

    2009-01-01

    In the PWR reactors, nickel based dissimilar welds have been presented susceptibilities the stress corrosion (S C). For the mitigation the problem a deposition of weld layers on the external surface of the nozzle is an alternative, viewing to provoke the compression of the region subjected to S C. This paper presents a preliminary study on the determination of welding parameters to obtain these welding overlayers. Welding depositions were performed on a test piece welded with nickel 182 alloy, simulating the conditions of a nozzle used in a PWR nuclear power plant. The welding process was the GTAW (Gas Tungsten Arc Welding), and a nickel 52 alloy as addition material. The overlayers were performed on the base metals, carbon steel an stainless steel, changing the welding parameters and verifying the the time of each weld filet. After that, the samples were micro structurally characterized. The macro structures and the microstructures obtained through optical microscopy and Vickers microhardness are presented. The preliminary results make evident the good weld quality. However, a small weld parameters influence used in the base material microstructure (carbon steel and stainless steel). The obtained results in this study will be used as reference in the construction of a mock up which will simulate all the conditions of a pressurizer nozzle of PWR reactor

  14. Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

    International Nuclear Information System (INIS)

    Chaves, Luiz C.; Curi, Marcos F.

    2017-01-01

    In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

  15. Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chaves, Luiz C.; Curi, Marcos F., E-mail: marcos.curi@cefet-rj.br [Centro Federal de Educação Tecnológica Celso Suckow da Fonseca (CEFET-RJ), Rio de Janeiro, RJ (Brazil). Department of Mechanical Engineering

    2017-07-01

    In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

  16. Secondary water chemistry control practices and results of the Japanese PWR plants

    International Nuclear Information System (INIS)

    Maeda, Akihiro; Shoda, Yasuhiko; Ishihara, Nobuo; Murata, Kazutoyo; Fujiwara, Hiroyuki; Hayakawa, Hitoshi; Matsuda, Tadashi

    2012-09-01

    In Japan, since the start of the operation of the first PWR plant, Mihama Unit-1 in 1970, 24 PWR plants have been built by 2010, and all of them are in operation. Due to the plant-specific needs of management, and by flexibly incorporating the state-of-the-art insights into the design, the system configurations of the plants vary so many as 15 types. Meanwhile, the geographical feature of Japan makes all the Japanese PWR plants to have condensers cooled by sea water, and all the plants have a common system with a full-flow Condensate Polisher System (CPS). To prevent corrosion, continued improvements of the secondary water chemistry management has been performed like other countries, and one of the major features of the Japanese PWR plants is an enhanced provision for the condenser leakage. The water quality of SG (Steam Generator) has been significantly improved by the provision for the sea water leakage, in combination with other improvements in water chemistry management. Also in Japan, almost all of the treatments of the spent polisher resin and the wastewater are performed within the power plant sites. To facilitate the treatment of the waste water and the regeneration of the spent resins, either ammonia or ETA (Ethanol Amine) is selected as the pH adjustment agent for the secondary system water. Also at the ammonia treatment, high pH accomplishes the inhibition of the piping wall thinning and the lower iron transportation into SGs. In addition, the iron transported into the SG is removed by the chemical conditioning treatment called ASCA (Advanced Scale Conditioning Agent). This provides the effective recovery of the SG heat-transfer performance, and the improved SG support plate BEC (Broached Egg Crate) hole blockage rates. Basically in Japan, the secondary water chemistry management has been improved based on a single basic specification, for the variety of the plant configurations, with the plant-specific investigations and analyses. This paper summarizes

  17. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  18. 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)

  19. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  20. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  1. Development of gadolinia bearing fuel for PWR

    International Nuclear Information System (INIS)

    Seki, Kazuichiro

    1986-01-01

    In the PWR power plants in Japan, the long-period operation cycle was extended legally to a maximum of 13 months from the conventional about 9 months in fiscal 1980. With this move, as a new type of fuel with burnable-poison-rod function, the development was started of gadolinia-bearing (gadolinium oxide) fuel, gadolinia being contained in the fuel pellets. The basic technology studies were completed in fiscal 1984. Actual irradiation of the fuel in Unit 2 of the Oi Power Station was then started in July 1984, demonstrating validity of the design. Meanwhile, the rapid power-up fest and the fuel center temperature measurement are conducted in an overseas reactor from fiscal 1983. The following are described: functions of the burnable absorber, the need for gadolinia-bearing fuel, experiences with gadolinia-bearing fuel, problems in the design and production of gadolinia-bearing fuel, the development of gadolinia-bearing fuel. (Mori, K.)

  2. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  3. 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.)

  4. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  5. Microprobe analysis and scanning electron microscope on PWR fuels

    International Nuclear Information System (INIS)

    Giannetto, B.

    1996-01-01

    In this text we present the apparatus used, in the CEA centers of Saclay and Cadarache, for analysis PWR spent fuels and we give results for Uranium oxide fuels and mixed oxide fuels. 5 figs., 26 photos

  6. 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.

  7. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  8. GO evaluation of a PWR spray system. Final report

    International Nuclear Information System (INIS)

    Long, W.T.

    1975-08-01

    GO is a reliability analysis methodology developed over the years from 1960 to the present by Kaman Sciences Corporation, Colorado Springs, Colorado. In this report the GO methodology is presented and its application demonstrated by performing a reliability analysis of a conceptual PWR Containment Spray System. Certain numerical results obtained are compared with those of a prior fault tree analysis of the same system as documented in the 11 January 1973 draft report, A Fault Tree Evaluation of a PWR Spray System

  9. Calculations of the neutron environment inside a PWR containment

    International Nuclear Information System (INIS)

    Hopkins, W.C.

    1979-01-01

    Neutron dose rates inside a PWR containment have been calculated using the DOT-DOMINO-MORSE technique. These dose rates are compared to measurements performed by teams from the Health Physics Division of the Oak Ridge National Laboratory and the Lawrence Livermore Laboratory. A simple method for extrapolating neutron dose rates at the top of the refueling pool to other areas in PWR containments is outlined

  10. Maintenance Technology and its Applications for PWR Plants

    International Nuclear Information System (INIS)

    Wachi, E.; Nishitani, J.; Okimura, K.; Tokunaga, H.

    2012-01-01

    Of the 24 PWR plants in Japan, eleven have been operated for more than 30 years. Accordingly, it has become extremely important to take measures against ageing structures and components in order to achieve safe and reliable long term operation of these plants. In this paper, a concept of the ageing countermeasure for PWR in Japan is outlined and then representative technologies related to various maintenance activities are presented. (author)

  11. Stress transfer to the Denali and other regional faults from the M 9.2 Alaska earthquake of 1964

    Science.gov (United States)

    Bufe, C.G.

    2004-01-01

    Stress transfer from the great 1964 Prince William Sound earthquake is modeled on the Denali fault, including the Denali-Totschunda fault segments that ruptured in 2002, and on other regional fault systems where M 7.5 and larger earthquakes have occurred since 1900. The results indicate that analysis of Coulomb stress transfer from the dominant earthquake in a region is a potentially powerful tool in assessing time-varying earthquake hazard. Modeled Coulomb stress increases on the northern Denali and Totschunda faults from the great 1964 earthquake coincide with zones that ruptured in the 2002 Denali fault earthquake, although stress on the Susitna Glacier thrust plane, where the 2002 event initiated, was decreased. A southeasterlytrending Coulomb stress transect along the right-lateral Totschunda-Fairweather-Queen Charlotte trend shows stress transfer from the 1964 event advancing slip on the Totschunda, Fairweather, and Queen Charlotte segments, including the southern Fairweather segment that ruptured in 1972. Stress transfer retarding right-lateral strike slip was observed from the southern part of the Totschunda fault to the northern end of the Fairweather fault (1958 rupture). This region encompasses a gap with shallow thrust faulting but with little evidence of strike-slip faulting connecting the segments to the northwest and southeast. Stress transfer toward failure was computed on the north-south trending right-lateral strike-slip faults in the Gulf of Alaska that ruptured in 1987 and 1988, with inhibitory stress changes at the northern end of the northernmost (1987) rupture. The northern Denali and Totschunda faults, including the zones that ruptured in the 2002 earthquakes, follow very closely (within 3%), for about 90??, an arc of a circle of radius 375 km. The center of this circle is within a few kilometers of the intersection at depth of the Patton Bay fault with the Alaskan megathrust. This inferred asperity edge may be the pole of counterclockwise

  12. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    International Nuclear Information System (INIS)

    Malet, J.; Mimouni, S.; Manzini, G.; Xiao, J.; Vyskocil, L.; Siccama, N.B.; Huhtanen, R.

    2015-01-01

    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

  13. Backbone dynamics of reduced plastocyanin from the cyanobacterium Anabaena variabilis: Regions involved in electron transfer have enhanced mobility

    DEFF Research Database (Denmark)

    Ma, L.X.; Hass, M.A.S.; Vierick, N.

    2003-01-01

    The dynamics of the backbone of the electron-transfer protein plastocyanin from the cyanobacterium Anabaena variabilis were determined from the N-15 and C-13(alpha) R-1 and R-2) relaxation rates and steady-state [H-1]-N-15 and [H-1]-C-13 nuclear Overhauser effects (NOEs) using the model......-free approach. The C-13 relaxation studies were performed using C-13 in natural abundance. Overall, it is found that the protein backbone is rigid. However, the regions that are important for the function of the protein show moderate mobility primarily on the microsecond to millisecond time scale. These regions...... are the "northern" hydrophobic site close to the metal site, the metal site itself, and the "eastern" face of the molecule. In particular, the mobility of the latter region is interesting in light of recent findings indicating that residues also on the eastern face of plastocyanins from prokaryotes are important...

  14. Accumulation, transfer, and environmental risk of soil mercury in a rapidly industrializing region of the Yangtze River Delta, China

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Biao; Yan, Lianxiang; Sun, Weixia; Zhao, Yongcun; Shi, Xuezheng [Chinese Academy of Sciences, Nanjing (China). State Key Lab. of Soil and Sustainable Agriculture; Wang, Mei [Chinese Academy of Sciences, Nanjing (China). State Key Lab. of Soil and Sustainable Agriculture; Graduate Univ. of the Chinese Academy of Sciences, Beijing (China); Weindorf, David C. [Louisiana State Univ., Baton Rouge, LA (United States). AgCenter

    2011-06-15

    Purpose: Mercury (Hg) accumulation and transfer in soil ecosystems has been altered on local, regional, and even global scales, and their environmental risk has increasingly been a concern to the public and the scientific community. Materials and methods: A county level region in Zhangjiagang County, the Yangtze River Delta (YRD) region of China and a factory with Hg-contaminated wastewater discharging within the region were selected to study the accumulation, bioavailability, and transfer of Hg from different sources in soils and crops under rapid industrialization, urbanization, and intensive agricultural activities. Regional soil samples close to and away from factories and local soil and crop samples around a typical factory were collected in the YRD region of China. Soil and crop Hg and basic soil properties were examined. Results and discussion: Significant soil Hg accumulation was found in soils away from factories regardless of Cambosols (Entisols) and Anthrosols (Inceptisols), while the mobile HCl-extractable Hg (HCl-Hg) were greater in soils closer to factories due to a decrease and increase in soil pH and organic matter. A high level of soil total Hg (T-Hg) was found around the factory, and soil and crop Hg accumulation in the vicinity of the factory was localized with an exponential decrease as distance away from the wastewater discharge outlet increased. Although Hg accumulated in these soils, the T-Hg levels at only a few sampling sites in acidic Anthrosols area were found to exceed the second most stringent critical value of Chinese Environmental Quality Standards for Soils. Conclusions: Considering the cessation of Hg-containing agrochemicals and limitation of effects of industrial activities on Hg accumulation, more attention should be paid to the changes in soil properties and crop rotations than controlling the pathways of Hg entering soils because the current environmental risk is mobilization of accumulated soil Hg. (orig.)

  15. Backbone dynamics of reduced plastocyanin from the cyanobacterium Anabaena variabilis: Regions involved in electron transfer have enhanced mobility

    DEFF Research Database (Denmark)

    Ma, L.X.; Hass, M.A.S.; Vierick, N.

    2003-01-01

    The dynamics of the backbone of the electron-transfer protein plastocyanin from the cyanobacterium Anabaena variabilis were determined from the N-15 and C-13(alpha) R-1 and R-2) relaxation rates and steady-state [H-1]-N-15 and [H-1]-C-13 nuclear Overhauser effects (NOEs) using the model-free appr......The dynamics of the backbone of the electron-transfer protein plastocyanin from the cyanobacterium Anabaena variabilis were determined from the N-15 and C-13(alpha) R-1 and R-2) relaxation rates and steady-state [H-1]-N-15 and [H-1]-C-13 nuclear Overhauser effects (NOEs) using the model...... are the "northern" hydrophobic site close to the metal site, the metal site itself, and the "eastern" face of the molecule. In particular, the mobility of the latter region is interesting in light of recent findings indicating that residues also on the eastern face of plastocyanins from prokaryotes are important...

  16. The post-Variscan development of the British Isles within a regional transfer zone influenced by orogenesis

    Science.gov (United States)

    Peacock, D. C. P.

    2004-12-01

    The break-up of Pangaea after the Variscan Orogeny included rifting extending southwards from the Barents Sea via the Norwegian-Greenland Rift and into the North Sea, and northwards from the Central Atlantic. These two major rift systems interacted to form an approximately 1200-km-wide transfer zone across the British Isles, where a complex network of basins developed during the Mesozoic. Fault patterns were commonly controlled by reactivation of Precambrian, Caledonian and Variscan structures. The two main rift systems were unable to breach this regional transfer zone, where the crust had been thickened by the Caledonian and Variscan orogenies, until the Eocene. Breaching did not occur down the North Sea and through the English Channel because of Alpine contraction in NW Europe. Instead, breaching occurred around the west of Ireland and NW Scotland, so the British Isles remained connected to Europe rather than to the North American Plate.

  17. An optimization model for collection, haul, transfer, treatment and disposal of infectious medical waste: Application to a Greek region.

    Science.gov (United States)

    Mantzaras, Gerasimos; Voudrias, Evangelos A

    2017-11-01

    simulation. The model was applied to the Region of East Macedonia - Thrace in Greece. The optimum solution resulted in one treatment plant located in the sanitary landfill area of Chrysoupolis, required no transfer stations and had a total management cost of 38,800 €/month or 809 €/t. If a treatment plant is sited in the most eastern part of the Region, i.e., the industrial area of Alexandroupolis, the optimum solution would result in a transfer station of 23 m 3 , located near Kavala General Hospital, and a total cost of 39,800 €/month or 831 €/t. A sensitivity analysis was conducted and two alternative scenarios were optimized. In the first scenario, a 15% rise in fuel cost and in the second scenario a 25% rise in IMW production were considered. At the end, a cost calculation in €/t/km for every type of vehicle used for haul and transfer was conducted. Also, the cost of the whole system was itemized and calculated in €/t/km and €/t. The results showed that the higher percentage of the total cost was due to the construction of the treatment plant. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Heat transfer and flow region characteristics study in a non-annular channel between rotor and stator

    Directory of Open Access Journals (Sweden)

    Nili-Ahmadabadi M.

    2012-01-01

    Full Text Available This paper will present the results of the experimental investigation of heat transfer in a non-annular channel between rotor and stator similar to a real generator. Numerous experiments and numerical studies have examined flow and heat transfer characteristics of a fluid in an annulus with a rotating inner cylinder. In the current study, turbulent flow region and heat transfer characteristics have been studied in the air gap between the rotor and stator of a generator. The test rig has been built in a way which shows a very good agreement with the geometry of a real generator. The boundary condition supplies a non-homogenous heat flux through the passing air channel. The experimental devices and data acquisition method are carefully described in the paper. Surface-mounted thermocouples are located on the both stator and rotor surfaces and one slip ring transfers the collected temperature from rotor to the instrument display. The rotational speed of rotor is fixed at three under: 300rpm, 900 rpm and 1500 rpm. Based on these speeds and hydraulic diameter of the air gap, the Reynolds number has been considered in the range: 4000transfer and pressure drop coefficients are deduced from the obtained data based on a theoretical investigation and are expressed as a formula containing effective Reynolds number. To confirm the results, a comparison is presented with Gazley’s (1985 data report. The presented method and established correlations can be applied to other electric machines having similar heat flow characteristics.

  19. HIGH COOLING WATER TEMPERATURE EFFECTS ON DESIGN AND OPERATIONAL SAFETY OF NPPS IN THE GULF REGION

    Directory of Open Access Journals (Sweden)

    BYUNG KOO KIM

    2013-12-01

    Full Text Available The Arabian Gulf region has one of the highest ocean temperatures, reaching above 35 degrees and ambient temperatures over 50 degrees in the summer. Two nuclear power plants (NPP are being introduced in the region for the first time, one at Bushehr (1,000 MWe PWR plant from Russia, and a much larger one at Barakah (4X1,400 MWe PWR from Korea. Both plants take seawater from the Gulf for condenser cooling, having to modify the secondary/tertiary side cooling systems design by increasing the heat transfer surface area from the country of origin. This paper analyses the secondary side of a typical PWR plant operating under the Rankine cycle with a simplified thermal-hydraulic model. Parametric study of ocean cooling temperatures is conducted to estimate thermal efficiency variations and its associated design changes for the secondary side. Operational safety is reviewed to deliver rated power output with acceptable safety margins in line with technical specifications, mainly in the auxiliary systems together with the cooling water temperature. Impact on the Gulf seawater as the ultimate heat sink is considered negligible, affecting only the adjacent water near the NPP site, when compared to the solar radiation on the sea surface.

  20. Henry Walter Bates: a traveler naturalist in the amazon region and the process of information transference

    OpenAIRE

    Ferreira, Rubens da Silva

    2004-01-01

    A partir do século XVII iniciou-se na Amazônia toda uma movimentação de viajantes/naturalistas atraídos pela biossociodiversidade dessa região dominada por uma floresta tropical. Henry Bates (1825-1892), estudioso de história natural, foi um deles, tendo, porém, se deslocado para o Norte do Brasil entre os anos de 1848 e 1859. Nesse contexto, o presente paper tem como objetivo analisar o processo de transferência das informações produzidas por esse viajante naturalista após 11 anos de trabalh...

  1. Effect of water chemistry on deposition for PWR plant operation

    International Nuclear Information System (INIS)

    Le Calvar, Marc; Bretelle, J. L.; Cailleaux, J. P.; Lacroix, R.; Guivarch, M.; Gay, N.; Taunier, S.; Gressier, F.; Varry, P.; Corredera, G.; Alos-Ramos, O.; Dijoux, M.

    2012-09-01

    For Pressurized Water Reactor (PWR) operation, water chemistry guidelines, specifications and associated surveillance programs are key to avoid deposition of oxides. Deposition of oxides can be detrimental by disrupting results of flow measurements, decreasing the thermal exchange capacity, or even by impairing safety. This paper describes the most important cases of deposition, their consequences for operation, and the implemented improvements to avoid their reoccurrence. Deposition that led to a Crud Induced Power Shift (CIPS) is also described. In the primary and in the secondary sides, orifice plates are typically used for measuring feedwater flow rate in nuclear power plants. Feedwater flow rates are used for control purposes and are important safety parameters as they are used to determine the plant's operating power level. Fouling of orifice plates in the primary side has been found during surveillance testing. For reactor coolant pumps, the formation of deposits on the seal No.1 can cause abnormally high or low leak rates through the seal. The leak rate through this seal must be carefully maintained within a prescribed range during plant operation. In the secondary side, orifice plate fouling has been the cause of feedwater flow/reference thermal power drift. For the steam generators (SG), magnetite deposition has led to fouling of the tube bundle, clogging of the quadri-foiled support plate holes and hard sludge formation on the base plate. For the generators, copper hollow conductors are widely used. Buildup of copper oxides on the interior walls of copper conductors has caused insufficient heat transfer. All these deposition cases have received adequate attention, understanding and response via improvement of our surveillance programs. (authors)

  2. Soil-to-plant transfer factors for natural radionuclides in the Brazilian cerrado region

    Energy Technology Data Exchange (ETDEWEB)

    Jacomino, Vanusa M.F.; Oliveira, Kerley A.; Menezes, Maria Angela de B., E-mail: vmfj@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Mello, Jaime de; Silva, David F. da, E-mail: jwvmello@ufv.b [Universidade Federal de Vicosa (UFV), MG (Brazil). Dept. de Solos; Siqueira, Maria C.; Taddei, Maria H.; Dias, Fabiana F., E-mail: mc_quimica@hotmail.co, E-mail: mhtaddei@cnen.gov.b, E-mail: fdias@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN-MG), Pocos de Caldas, MG (Brazil). Lab. de Pocos de Caldas (LAPOC)

    2009-07-01

    Large amounts of phosphogypsum produced have been attracting attention of Radiological Protection institutions and Environmental Protection agencies worldwide, given its high potential for environmental contamination. In Brazil, this material has been used for several decades, especially for agricultural purposes. Due to the presence of radionuclides in its composition, it is necessary to understand the mechanisms for natural radionuclide transfer in the soil/plant system and to evaluate if the use of phosphogypsum in soil contributes to increased exposition of humans to natural radioactivity. Experiments were accomplished in a greenhouse with lettuce cultivation in two types of soil (sandy and clayey) fertilized with four different amounts of phosphogypsum. Samples of phosphogypsum, soil, lettuce and drainage water were then analyzed for key radionuclides. {sup 238}U and {sup 232}Th analyses were carried out by Neutron Activation Analysis; {sup 226}Ra, {sup 228}Ra, and {sup 210}Pb by analyzed by Gamma Spectrometry; and {sup 210}Po by Alpha Spectrometry Technique. Finally, Transfer Factors of soil-plant were calculated as well as annual contribution to the effective dose due to the ingestion of lettuces. {sup 22}'6Ra average specific activity in phosphogypsum samples (252 Bq kg{sup -1}) was below the maximum level recommended by USEPA, which is 370 Bq.kg{sup -1} for agricultural use. Although most of the results for mean specific activity of radionuclides in lettuce presented values below the Minimum Detectable Activity (MDA), Transfer Factors were estimated for those conditions in which the mean specific activity proved to be superior to MDA. Values ranged from 1.8 10{sup -3} to 2.3 10{sup -2} for {sup 232}Th; 3.5 10{sup -}'2 to 4.1 10{sup -2} for {sup 226}Ra, 2.4 10{sup -1} to 3.2 10{sup -}'1 for {sup 228}Ra, and 3.5 10{sup -2} to 8.5 10{sup -2} for {sup 210}Po, depending on the type of soil used for planting vegetables. In general, results

  3. EMERALD, Radiation Release and Dose after PWR Accident for Design Analysis and Operation Analysis

    International Nuclear Information System (INIS)

    Brunot, W.K.; Fray, R.R.; Gillespie, S.G.

    1988-01-01

    1 - Description of problem or function: The EMERALD program is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large pressurized water reactor (PWR). The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place in the plant. For example, in the calculation of the doses resulting from a loss-of-coolant accident the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program. Subroutines are also included which calculate the on-site and off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the twenty-five isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials during abnormal operation of a PWR. These include design, operational, and licensing studies. 2 - Method of solution - Explicit solutions of first-order linear differential equations are included. In addition, a subroutine is provided which solves a set of simultaneous linear algebraic equations. 3 - Restrictions on the complexity of the problem - Maxima of: 25 isotopes, 7 time periods, 15 volumes or components, 10

  4. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  5. CORD-2 package for PWR design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.

    1994-01-01

    The CORD-2 package is designed to provide a modern, independent calculational tool for reactor core calculations. It provides options that are essential for modeling the advanced features of fuel assemblies. Its development is part of a wider effort to establish country's own expertise in nuclear design and safety analysis. The package provides not only the calculational modules, but also the data management support facilities. It has been implemented on VAX/VMS and on PC/DOS, but extension to other systems is quite straightforward. The main components and the calculational methods are briefly described. The results of the validation programme are presented. They include the comparison of the calculated results with the measured values of ten cycles of the Krsko nuclear power plant and for the IAEA test case Almaraz, with special emphasis on the first cores at hot-zero power conditions. The results of the validation programme shows that CORD-2 is applicable for design level PWR core calculations. (authors). 9 refs., 4 figs., 3 tabs

  6. Alloy development for high burnup cladding (PWR)

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, K. H.; Kim, S. J.; Choi, B. K.; Kim, J.M.

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs

  7. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  8. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  9. Minimization of radioactive liquids released from PWR

    International Nuclear Information System (INIS)

    Yoshikawa, Hideo; Kohri, Masaharu.

    1981-01-01

    The quantity of radioactive substances in the liquids released into the environment from a PWR power station in normal operation was determined, following the path from the sources of generation, that is, the equipments in primary and secondary cooling systems, to the release into the environment after the radioactive substances were removed in treatment facilities. The quantity of radioactive substances released from primary and secondary systems was determined for each source of generation in a standard plant, and the results were examined. As the concrete example of reducing the release on the basis of ''As low as reasonably achievable'' concept, the increase of letdown flow rate and the installation of a condensate-desalting column are reported. As the sources of generation, the primary coolant formed by shim bleed and the drain from primary system equipments, the drain from an auxiliary building floor, radiochemistry waste solution and the drain from intermediate cooling system, the waste water of washing and shower bath, the drain from a turbine building floor, and the blow-down waste from steam generators are enumerated. The concentration of radioactive substances in primary and secondary coolants, the decontamination factor of waste treatment equipments and the measures for reducing the release are described. (Kako, I.)

  10. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    Vaidian, T.A.; Karmakar, G.; Rajagopal, R.; Shankar, V.; Patil, R.K.

    1994-01-01

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  11. 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.

  12. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    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

  13. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  14. Electroproduction of η mesons in the S11(1535) resonance region at high momentum transfer

    International Nuclear Information System (INIS)

    Dalton, M. M.; Adams, G. S.; Moziak, B.; Stoler, P.; Villano, A.; Ahmidouch, A.; Danagoulian, S.; Angelescu, T.; Malace, S.; Arrington, J.; Hafidi, K.; Holt, R. J.; Reimer, P. E.; Schulte, E.; Zheng, X.; Asaturyan, R.; Mkrtchyan, H.; Navasardyan, T.; Tadevosyan, V.; Baker, O. K.

    2009-01-01

    The differential cross section for the process p(e,e ' p)η has been measured at Q 2 ∼5.7 and 7.0(GeV/c) 2 for center-of-mass energies from threshold to 1.8 GeV, encompassing the S 11 (1535) resonance, which dominates the channel. This is the highest momentum-transfer measurement of this exclusive process to date. The helicity-conserving transition amplitude A 1/2 , for the production of the S 11 (1535) resonance, is extracted from the data. Within the limited Q 2 now measured, this quantity appears to begin scaling as Q -3 --a predicted, but not definitive, signal of the dominance of perturbative QCD at Q 2 ∼5 (GeV/c) 2 .

  15. Numerical study of heat transfer and combustion in IC engine with a porous media piston region

    International Nuclear Information System (INIS)

    Zhou, Lei; Xie, Mao-Zhao; Luo, Kai Hong

    2014-01-01

    Based on superadiabatic combustion in porous medium (PM), the porous medium engine as a new combustion concept is proposed to achieve high combustion efficiency and low emissions. In this paper, an axisymmetric model with detailed chemistry and two-temperature treatment is implemented into a variant of the KIVA-3V code to simulate the working process of the PM engine. Comparisons with the same engine but without PM are conducted. Temperature evolution of the PM and its effects are discussed in detail. Key factors affecting heat transfer, combustion and emissions of the PM engine, such as porosity, the initial PM temperature and equivalence ratio, are analyzed. The results show that the characteristics of heat transfer, emissions and combustion of the PM engine are superior to the engine without PM, providing valuable support for the PM engine concept. In particular, the PM engine is shown to sustain ultra lean combustion. - Graphical abstract: In the PM engine, a PM reactor is mounted on the piston head as shown in Fig. 1 which shows the schematic diagram of the computational domain. The heat exchange process between PM material and compressed air increases with upward motion of piston at compression stroke. At the TDC, almost all the air is compressed and closed to PM volume, meanwhile, the fuel is injected into PM chamber to achieve homogenization combustion. - Highlights: •Two-temperature treatment studies the working process of the PM engine. •Self-balancing temperature of the PM determines the continued and stable work. •Stronger heat exchange occurs between gas and PM with smaller porosity. •The PM engine can have lower levels of NO x , unburnt HC and CO emissions

  16. Extending amulti-scale parameter regionalization (MPR) method by introducing parameter constrained optimization and flexible transfer functions

    Science.gov (United States)

    Klotz, Daniel; Herrnegger, Mathew; Schulz, Karsten

    2015-04-01

    A multi-scale parameter-estimation method, as presented by Samaniego et al. (2010), is implemented and extended for the conceptual hydrological model COSERO. COSERO is a HBV-type model that is specialized for alpine-environments, but has been applied over a wide range of basins all over the world (see: Kling et al., 2014 for an overview). Within the methodology available small-scale information (DEM, soil texture, land cover, etc.) is used to estimate the coarse-scale model parameters by applying a set of transfer-functions (TFs) and subsequent averaging methods, whereby only TF hyper-parameters are optimized against available observations (e.g. runoff data). The parameter regionalisation approach was extended in order to allow for a more meta-heuristical handling of the transfer-functions. The two main novelties are: 1. An explicit introduction of constrains into parameter estimation scheme: The constraint scheme replaces invalid parts of the transfer-function-solution space with valid solutions. It is inspired by applications in evolutionary algorithms and related to the combination of learning and evolution. This allows the consideration of physical and numerical constraints as well as the incorporation of a priori modeller-experience into the parameter estimation. 2. Spline-based transfer-functions: Spline-based functions enable arbitrary forms of transfer-functions: This is of importance since in many cases the general relationship between sub-grid information and parameters are known, but not the form of the transfer-function itself. The contribution presents the results and experiences with the adopted method and the introduced extensions. Simulation are performed for the pre-alpine/alpine Traisen catchment in Lower Austria. References: Samaniego, L., Kumar, R., Attinger, S. (2010): Multiscale parameter regionalization of a grid-based hydrologic model at the mesoscale, Water Resour. Res., doi: 10.1029/2008WR007327 Kling, H., Stanzel, P., Fuchs, M., and

  17. Heat transfer in the inner and boundary region of pebble beds

    International Nuclear Information System (INIS)

    Robold, K.

    1982-07-01

    The effective thermal conductivity in the inner and boundary region of pebble beds have been measured. The experiments were carried out in evacuated pebble beds and beds with stagnant Helium (p = 700...850 mbar). The temperature range was 300 to 1900 K. The experimental results are described by new models. (orig.) [de

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

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    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 −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

  19. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    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

  20. Numerical simulation of turbulent unsteady compressible pipe flow with heat transfer in the entrance region

    Science.gov (United States)

    Ziaei-Rad, Masoud; Nouri-Broujerdi, Ali

    2008-12-01

    In this paper, the compressible gas flow through a pipe subjected to wall heat flux in unsteady condition in the entrance region is investigated numerically. The coupled conservation equations governing turbulent compressible viscous flow in the developing region of a pipe are solved numerically under different thermal boundary conditions. The numerical procedure is a finite-volume-based finite-element method applied to unstructured grids. The convection terms are discretized by the well-defined Roe method, whereas the diffusion terms are discretized by a Galerkin finite-element formulation. The temporal terms are evaluated based on an explicit fourth-order Runge-Kutta scheme. The effect of different thermal conditions on the pressure loss of unsteady flow is investigated. The results show that increase in the inflow temperature or pipe-wall heat flux increases the pressure drop or decreases the mass flow rate in the pipe.

  1. Asian regional co-operative project on food irradiation: Technology transfer

    International Nuclear Information System (INIS)

    1992-01-01

    These Proceedings include the final reports of work performed by different institutions under the scope of Phase II of the Asian Regional Co-operative Project on Food Irradiation. The topics covered include the disinfestation and decontamination of stored products; improvements in the hygiene of processed seafood; insect disinfestation of fruits; and sprout inhibition of root crops. The individual presentations are indexed separately. Refs, figs and tabs

  2. Proton-transfer reaction mass spectrometry (PTR-MS) for the authentication of regionally unique South African lamb.

    Science.gov (United States)

    Erasmus, Sara W; Muller, Magdalena; Alewijn, Martin; Koot, Alex H; van Ruth, Saskia M; Hoffman, Louwrens C

    2017-10-15

    The volatile fingerprints of South African lamb meat and fat were measured by proton-transfer mass spectrometry (PTR-MS) to evaluate it as an authentication tool. Meat and fat of the Longissimus lumborum (LL) of lambs from six different regions were assessed. Analysis showed that the volatile fingerprints were affected by the origin of the meat. The classification of the origin of the lamb was achieved by examining the calculated and recorded fingerprints in combination with chemometrics. Four different partial least squares discriminant analysis (PLS-DA) models were fitted to the data to classify lamb meat and fat samples into "region of origin" (six different regions) and "origin" (Karoo vs. Non-Karoo). The estimation models classified samples 100% correctly. Validation of the first two models gave 42% (fat) and 58% (meat) correct classification of region, while the second two models performed better with 92% (fat) and 83% (meat) correct classification of origin. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. Electroproduction of eta Mesons in the S11(1535) Resonance Region at High Momentum Transfer

    International Nuclear Information System (INIS)

    Dalton, Mark

    2008-01-01

    The differential cross-section for the exclusive process p(e, e(prime)p)η has been measured at Q2 ∼ 5.7 and 7.0 (GeV/c)2, which represents the highest momentum transfer measurement of this to date, significantly higher than the previous highest at Q2 ∼ 3.6 (GeV/c)2. Data was taken for centre-of-mass energies from threshold to ∼1.8 GeV, encompassing the S11(1535) resonance, which dominates the pη channel. The total cross section is obtained, from which is extracted the helicity-conserving transition amplitude A1/2, for the production of the S11(1535) resonance. This quantity appears to begin scaling as Q-3, a predicted signal of the dominance of perturbative QCD, within the Q2 range of this measurement. No currently available theoretical predictions can account for the behaviour of this quantity over the full measured range of Q2.

  4. Application of the MELCOR code to design basis PWR large dry containment analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  5. Nupec thermal hydraulic test to evaluate post-DNB characteristics for PWR fuel assemblies (1. general test plan and results)

    International Nuclear Information System (INIS)

    Norio, Kono; Kenji, Murai; Kaichiro, Misima; Takayuki, Suemura; Yoshiei, Akiyama; Keiichi, Hori

    2001-01-01

    In the present thermal hydraulic design of Pressurized Water Reactor (PWR), a departure from nucleate boiling (DNB) under anticipated transient conditions is not allowed. However, it is recognized that the DNB dose not cause a fuel rod failure immediately, and a suitable reactor trip can prevent the core from severe damages. If the fuel rod temperature under the post-DNB conditions can be accurately evaluated, the potentially existing margin in the present design method will be quantitatively assessed. To establish the heat transfer evaluation method on post-DNB event for PWR thermal hydraulic design, Nuclear Power Engineering Corporation (NUPEC) started a program, NUPEC Thermal Hydraulic Test to Evaluate Post-DNB Characteristics for PWR Fuel Assemblies (NUPEC-TH-P), in 1995 (hereinafter the year means fiscal year) under the sponsorship of Ministry of Economy, Trade and industry (METI). This program is now under going until 2001. This paper is to show the overall plan and the status of NUPEC-TH-P. (authors)

  6. 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)

  7. Improved emergency elevated air release for simplified PWR

    International Nuclear Information System (INIS)

    Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

    1992-01-01

    In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

  8. Industry-wide survey of organics in PWR's

    International Nuclear Information System (INIS)

    Byers, W.A.; Richards, J.E.; Hobart, S.A.

    1986-01-01

    Interest in organic impurities found in Pressurized Water Reactors (PWR's) has stemmed from several sources. The most serious concern is that organic acids will increase cation conductivity, a parameter that is used to control power plant chemistry. This effect can complicate secondary water monitoring and control. Organics may foul or exhaust makeup demineralizers and condensate polishers, and thus result in increased operating costs or the in leakage of potentially corrosive agents into the steam generators. Some organics, however, such as mopholine and cyclohexylamine may reduce corrosion through oxygen scavenging or surface filming reactions, and may have a positive influence on the pH in areas of local corrosion. At the time this survey began, little information was available on the types or levels of organic impurities that are typically found in PWR's. this survey is intended to provide baseline data for future corrosion testing and to provide fundamental information that will be helpful in refining PWR chemistry guidelines and operating practices

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

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

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

  10. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  11. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  12. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

  13. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended 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. 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. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  14. Generation and transfer of internal variability in a regional climate model

    Directory of Open Access Journals (Sweden)

    Thorsten Simon

    2013-12-01

    Full Text Available There is a strong need for tools allowing the comparison between the performance of a regional climate model (RCM and the corresponding model providing lateral boundary conditions (LBC for the RCM, which is a global general circulation model (GCM in most cases. A method is presented to investigate the temporal scales on which a RCM is able to generate internal variability on its own and on which variability is copied from the driving model. This is implemented by a cross-spectral analysis between the RCM output and a bi-linearly interpolated version of the driving model, leading to an estimate of the coherence spectrum. Applying the aforementioned technique to surface temperature and temperature and specific humidity at 850 hPa from the RCM COSMO-CLM East Asia with a horizontal resolution of 50 km and its driving model ECHAM5, it was found that features in the spatial distribution of coherence are related to atmospheric dynamics in East Asia, e.g. monsoons and inter-tropical convergence zone (ITCZ. A further application to a double-nesting approach, where COSMO-CLM East Asia is the driving model for two domains – namely the Haihe catchment and the Poyang catchment – each with a horizontal resolution of 7 km, shows that the frequencies on which internal variability is generated by the driven model are much higher compared to the first nesting step. Concluding RCMs can produce a considerable variability on the respective temporal scales. This implies that a dynamical downscaling with a re-analysis as LBC is conceptually different to a regional re-analysis, i.e. data assimilation on the regional scale.

  15. Load-following operation of PWR plants

    International Nuclear Information System (INIS)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom's N4 plant, KWU's plants, ABB-CE's Systems 80+, and Westinghouse's AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author)

  16. Experimental study of film boiling heat transfer in steam-water two-phase flow

    International Nuclear Information System (INIS)

    Iwamura, Takamichi

    1986-05-01

    A steady-state film boiling experiment at void fractions between 0.6 and 0.95 was performed to investigate the film boiling heat transfer coefficient in dispersed flow and transition regions during the reflood phase of a PWR-LOCA. The film boiling heat transfer in these regions was assumed to be superimposed by three different mechanisms; radiation, forced convection to steam and droplet impingement on wall. The radiation and forced convection heat transfer coefficients were evaluated by using the Stefan-Boltzmann equation and the Dittus-Boelter equation, respectively. The thermodynamic non-equilibrium was taken into account in the forced convection heat transfer mode. A new correlation for the heat transfer coefficient due to droplet impingement was derived from the dispersed flow heat transfer model developed by Forslund and Rohsenow. The correlation is a function of steam and water velocities, void fraction, fluid properties and wall superheat. The agreement between calculated and experimentally derived heat transfer coefficients was fairly good for the present experiment. (author)

  17. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  18. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    Bayley, B.; Stilwell, W.E.; Kent, N.A.

    2004-01-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  19. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

  20. Structures and Materials of Reactor Internals for PWR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Kim, W. S.; Kwon, S. C.; Kwon, J. H.; Kim, Y. S.; Kim, H. P.; Yoo, C. S.; Lee, S. R.; Jung, M. K.; Hwang, S. S

    2007-10-15

    Nuclear reactor types in Korea are PWR type reactor (Westinghouse, Combustion Engineering, Farmatome type) and CANDU type reactor. Structures and Materials for reactor internal of PWR type were investigated. Reactor internal was composed of lower core support structure, upper core support assembly, incore instrumentation support structure. Lower core support structure of these structures is the most important. The major material for the reactor internal is type 304 and 316 stainless steel and radial support clevis bolts are made of Inconel. The main damage mechanism for reactor internal was IASCC and the effect of IASCC on reactor internal was investigated. The accident for reactor internal was also investigate.

  1. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  2. 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.

  3. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  4. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  5. Regional accumulation characteristics of cadmium in vegetables: Influencing factors, transfer model and indication of soil threshold content.

    Science.gov (United States)

    Yang, Yang; Chen, Weiping; Wang, Meie; Peng, Chi

    2016-12-01

    A regional investigation in the Youxian prefecture, southern China, was conducted to analyze the impact of environmental factors including soil properties and irrigation in conjunction with the use of fertilizers on the accumulation of Cd in vegetables. The Cd transfer potential from soil to vegetable was provided by the plant uptake factor (PUF), which varied by three orders of magnitude and was described by a Gaussian distribution model. The soil pH, content of soil organic matter (SOM), concentrations of Zn in the soil, pH of irrigation water and nitrogenous fertilizers contributed significantly to the PUF variations. A path model analysis, however, revealed the principal control of the PUF values resulted from the soil pH, soil Zn concentrations and SOM. Transfer functions were developed using the total soil Cd concentrations, soil pH, and SOM. They explained 56% of the variance for all samples irrespective of the vegetable genotypes. The transfer functions predicted the probability of exceeding China food safety standard concentrations for Cd in four major consumable vegetables under different soil conditions. Poor production practices in the study area involved usage of soil with pH values ≤ 5.5, especially for the cultivation of Raphanus sativus L., even with soil Cd concentrations below the China soil quality standard. We found the soil standard Cd concentrations for cultivating vegetables was not strict enough for strongly acidic (pH ≤ 5.5) and SOM-poor (SOM ≤ 10 g kg -1 ) soils present in southern China. It is thus necessary to address the effect of environmental variables to generate a suitable Cd threshold for cultivated soils. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. Os programas de transferência de renda e o voto regional nas eleições presidenciais de 2010

    Directory of Open Access Journals (Sweden)

    Maria Teresa Miceli Kerbauy

    2011-11-01

    Full Text Available A analise das eleições presidenciais de 2010 apontaram para um eleitorado cujas escolhas se diferenciam por regiões, seguindo o mapa da desigualdade do Brasil. A pauta da campanha esteve ligada à continuidade do governo Lula, especialmente à expansão da cobertura dos programas de transferência de renda e ampliação do consumo, incentivando a discussão sobre o papel desses programas na decisão do voto. Nesse, sentido, o artigo investiga a relação entre conhecimento e participação nos programas de redistribuição de renda, o impacto na percepção política do eleitorado e no voto e sua relação com a questão regional. Para esta análise utilizamos os dados do survey eleitoral ESEB realizado após as eleições de 2010.The analysis of the 2010 Presidential elections shows a divided electorate by regions, following the Brazilian map of inequality. The main campaign issues were the continuity of President Lula and the social programs. This article investigates the relationship between social programs and voter decision, and the role of region as a variable of explanation. The analysis is based on the CSES-ESEB2010 data.

  7. [Distribution and transferring of carbon in kast soil system of peak forest depression in humid subtropical region].

    Science.gov (United States)

    Pan, G; Sun, Y; Teng, Y; Tao, Y; Han, F

    2000-02-01

    Taking Guilin Yaji Karst Experiment Site as an exemple and with the methods of field monitoring and laboratory analysis, this paper studied the distribution and transferring of carbon in the karst soil system of peak forest depression in the humid subtropical region of China. The carbon pools in biomass, litters and soil organic matter(SOM) and their mobility as expressed by oxidizability and decomposition rate of SOM, the concentration of soil CO2 and the emission rate of CO2 from soil were investigated. The mobile carbon pool in the system supplied a rich source of CO2, which drived the karst process. When active karst process happened in Spring and Summer, over 60% of carbon in the output water was derived from soil CO2, as traced by delta 13 C distribution in the system. Therefore, owing to the carbon transfer in the pathway of air-plant-soil-water, karst process took place rather under soil-rock-water interface than under air-rock-water interface. Thus, the epigenetic karst process was driven and accelerated by soil as an interface of carbon environmental geochemistry.

  8. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  9. Carbon Capture and Storage in the Permian Basin, a Regional Technology Transfer and Training Program

    Energy Technology Data Exchange (ETDEWEB)

    Rychel, Dwight [Petroleum Tech Transfer Council, Oak Hill, VA (United States)

    2013-09-30

    The Permian Basin Carbon Capture, Utilization and Storage (CCUS) Training Center was one of seven regional centers formed in 2009 under the American Recovery and Reinvestment Act of 2009 and managed by the Department of Energy. Based in the Permian Basin, it is focused on the utilization of CO2 Enhanced Oil Recovery (EOR) projects for the long term storage of CO2 while producing a domestic oil and revenue stream. It delivers training to students, oil and gas professionals, regulators, environmental and academia through a robust web site, newsletter, tech alerts, webinars, self-paced online courses, one day workshops, and two day high level forums. While course material prominently features all aspects of the capture, transportation and EOR utilization of CO2, the audience focus is represented by its high level forums where selected graduate students with an interest in CCUS interact with Industry experts and in-house workshops for the regulatory community.

  10. Transfer reactions and multiple Coulomb excitation in the $^{100}$Sn Region

    CERN Multimedia

    It is proposed to continue our REX-ISOLDE program in the $^{100}$Sn region at HIE-ISOLDE at ~5 MeV/u. Earlier measurements, with a precision of 10-20%, at 3 MeV/u with REX-ISOLDE point to a deviation between the measured B(E2) values for the first excited 2$^{+}$ states in $^{110,108,106}$Sn compared to theoretical predictions. In addition, the trend of B(E2) values for the lighter isotopes, in particular $^{106}$Sn, appear to differ between low- and high-energy measurements. In line with our letter-of-intent we aim in a first step to address the electromagnetic properties of the first 2$^{+}$and 4$^{+}$ states in $^{110,108,106}$Sn using Coulomb excitation. In these measurements we will directly access the lifetimes of the first excited 4$^{+}$ states in $^{110,108,106}$Sn for the first time. The yield of $^{104}$Sn from the LaC$_{x}$ target will be revisited to clarify if the new solid state RILIS gives sufficient yield to expand the measurements to this isotope. Following this proposal we plan similar meas...

  11. 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

  12. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    Jaime, Guilherme D.G.; Oliveira, Mauro V.

    2011-01-01

    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

  13. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    Lelong, Franck

    2010-01-01

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  14. Directives and general design requirements for a small PWR

    International Nuclear Information System (INIS)

    Arrieta, L.A.

    1992-08-01

    A proposal of directives and general requirements for the development of a small PWR conceptual design is presented. These directives address the main safety, performance and economic design aspects. The purpose is to use this work as a base for a wide discussion, involving all project participants, culminating with the definition of the final directives and general requirements. (author)

  15. Seismic analysis of the core of a PWR reactor

    International Nuclear Information System (INIS)

    Preumont, A.

    1981-01-01

    The author develops successively: - a method for the generation of accelerograms compatible with the response spectrum; a model for the analysis of lateral deformations of the core of a PWR reactor under seismic excitation; a simple dynamic model of the fuel assembly including a vibration model. (MD)

  16. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  17. Reactor core design calculations and fuel management in PWR

    International Nuclear Information System (INIS)

    Ravnik, M.

    1987-01-01

    Computer programs and methods developed at J. Stefan Institute for nuclear core design of Krsko NPP are treated. development, scope, verification and organisation of core design procedure are presented. The core design procedure is applicable to any NPP of PWR type. (author)

  18. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits [fr

  19. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  20. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  1. Exhaust air cleaning of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Wang Rufan.

    1987-01-01

    This paper describes requirements, design criteria and major equipments of exhaust air cleaning for PWR nuclear power plants. The particularity of exhaust air cleaning for NPP is stressed in this paper. Finally, the exhaust air cleaning systems of Qinshan NPP are briefly introduced

  2. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  3. VHTR, ADS, and PWR spent nuclear fuel analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salome, J.A.D.; Cardoso, F.; Velasquez, C.E.; Pereira, F.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP: 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Rio de Janeiro (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear - CNEN, Rua General Severiano 82, Botafogo, Rio de Janeiro, RJ, CEP: 22290-040 (Brazil)

    2016-07-01

    The aim of this study is to analyze and compare the discharged-spent fuel of 3 types of nuclear systems: a Very High-Temperature Gas Reactor (VHTR), a lead-cooled Accelerator-Driven System (ADS) and a standard Pressurized Water Reactor (PWR). The two first systems, VHTR, and ADS were designed to use reprocessed fuels. UREX+ and GANEX techniques were used for the reprocessing processes respectively. The fuel burnup simulated for the systems in other works have been used to obtain the final composition of the spent fuel discharged. After discharge, the radioactivity, the radiotoxicity, and the decay heat were evaluated through the ORIGEN 2.1 code until 10{sup 7} years and compared to the literature. The spent nuclear waste (SNF) coming from reprocessing techniques and burned up in advanced reactors show that the radiotoxicity decreases below a conventional SNF from a typical PWR for the time studied. The VHTR and ADs have higher values of radioactivity, radiotoxicity and decay heat, because of the greater concentrations of plutonium and curium in these reactors than in the PWR. Fission products have the greatest contribution for the first 25 years over the parameters studied for a PWR. The most harmful fission products are: Ba{sup 137m}, Tc{sup 99}, I{sup 129} and Nb{sup 93m} and for actinides is the plutonium and curium.

  4. On the comparability of knowledge transfer activities – a case study at the German Baltic Sea Coast focusing regional climate services

    Directory of Open Access Journals (Sweden)

    I. Meinke

    2017-06-01

    Full Text Available In this article the comparability of knowledge transfer activities is discussed by accounting for external impacts. It is shown that factors which are neither part of the knowledge transfer activity nor part of the participating institution may have significant impact on the potential usefulness of knowledge transfer activities. Differences in the potential usefulness are leading to different initial conditions of the knowledge transfer activities. This needs to be taken into account when comparing different knowledge transfer activities, e.g., in program evaluations. This study is focusing on regional climate services at the German Baltic Sea coast. It is based on two surveys and experiences with two identical web tools applied on two regions with different spatial coverage. The results show that comparability among science based knowledge transfer activities is strongly limited through several external impacts. The potential usefulness and thus the initial condition of a particular knowledge transfer activity strongly depends on (1 the perceived priority of the focused topic, (2 the used information channels, (3 the conformity between the research agenda of service providing institutions and information demands in the public, as well as (4 on the spatial coverage of a service. It is suggested to account for the described external impacts for evaluations of knowledge transfer activities. The results show that the comparability of knowledge transfer activities is limited and challenge the adequacy of quantitative measures in this context. Moreover, as shown in this case study, in particular regional climate services should be individually evaluated on a long term perspective, by potential user groups and/or by its real users. It is further suggested that evaluation criteria should be co-developed with these stakeholder groups.

  5. A survey of blockage measurement methods used in PWR multi-rod experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E.D.; Jones, C.; Whitty, S. (AEA Reactor Services, Springfield (UK))

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author).

  6. A survey of blockage measurement methods used in PWR multi-rod experiments

    International Nuclear Information System (INIS)

    Hindle, E.D.; Jones, C.; Whitty, S.

    1986-05-01

    The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author)

  7. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  8. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

    International Nuclear Information System (INIS)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

    2011-01-01

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  9. Development of intelligent Eddy Current Testing (ECT) system for PWR steam generator tube inspection

    International Nuclear Information System (INIS)

    Kawata, K.; Kawase, N.; Kurokawa, M.; Asada, Y.

    2005-01-01

    The intelligent ECT system was developed for the inspection of heat transfer tubes of the steam generator of the PWR plant. It consists of intelligent probe, data acquisition unit and data analysis system. The probe combines 24 channels inclined lay out drive coils and thin film pick-up coils with built-in electric circuits to provide high inspection capability equivalent to rotating coil ECT and high-speed inspection equivalent to conventional bobbin coil ECT. The advanced data analysis system that has filtering and automatic analysis functions is also developed to enable fast and precise analysis of large volume inspection data. The system was qualified by confirmation tests in FY 2003 to show thinned thickness sizing accuracy within ± 5%. (T. Tanaka)

  10. Differential language expertise related to white matter architecture in regions subserving sensory-motor coupling, articulation, and interhemispheric transfer.

    Science.gov (United States)

    Elmer, Stefan; Hänggi, Jürgen; Meyer, Martin; Jäncke, Lutz

    2011-12-01

    The technique of diffusion tensor imaging (DTI) has been used to investigate alterations in white matter architecture following long-term training and expertise. Professional simultaneous interpreters (SI) provide an ideal model for the investigation of training-induced plasticity due to the high demands placed on sound to motor mapping mechanisms, which are vital for executing fast interpretations. In line with our hypothesis, we found clusters with decreased fractional anisotropy (FA) in the SI group in brain regions previously shown to support sensory-motor coupling mechanisms and speech articulation (cluster extent family-wise error corrected, P architecture indicated by lower FA values in the SI group in the most anterior and posterior parts of the corpus callosum. Our results suggest that language expertise is accompanied by plastic adaptations in regions strongly involved in motor aspects of speech and in interhemispheric information transfer. These results have implications for our understanding of language expertise in relation to white matter adaptations. Copyright © 2010 Wiley Periodicals, Inc.

  11. Colloids in PWR primary and secondary coolant. Innovative analytical methods

    International Nuclear Information System (INIS)

    Nowotka, Karsten; Guillodo, Michael; Burchardt, Carsten; Geier, Roland; Lehr, Robert; Stellwag, Bernhard

    2014-01-01

    Transport and deposition of corrosion products in the colloid size range between 1 nm and 1 μm are important for heat transfer performance and corrosion in primary and secondary cooling circuits of LWRs. Direct analysis of the properties of small-sized colloids (< 0.45 μm) is difficult due to the pronounced change of the physicochemical properties of coolant samples in sampling lines. An innovative method, based on a filter cascade and developed in the AREVA Technical Center, named 'Colloid Catcher' (CC, patent pending), permits on-line measurements of the properties of corrosion products in the coolant of LWRs. CC measurements are complementary to classic trace analysis addressing the soluble content. Low and high temperature (up to 330°C) test sections are available, depending on our customer's needs. The CC contains differential pressure detectors at each of the three consecutive membrane filters which allow for an in-situ characterization without modification of the corrosion products chemical nature due to temperature changes and subsequent exposure to the atmosphere. With this method, a 'Colloid Fingerprint' of the test solution can be obtained, ideal for an assessment of the transport and deposition of corrosion products in laboratory and on-site studies. The on-line data can of course be complemented by post filtration membrane characterization by digestion and/or optical methods. The high temperature CC serves at the same time as a sampling point for grab samples, with good reproducibility thanks to continuous liquid flow. The CC has been designed to be deployable on laboratory or industrial cooling circuits. The CC test sections have been qualified using AREVA Technical Center's test loops. First test results obtained with the LT CC are presented. Laboratory data can be used to back up existing results and data of on-site measurement campaigns at BWR and PWR plants which were determined with a basic version of the LT CC

  12. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code

    International Nuclear Information System (INIS)

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A.

    2005-01-01

    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)

  13. 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.

  14. Analysis of the return to power scenario following a LBLOCA in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

  15. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  16. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  17. Transferability of microsatellite markers in Syagrus coronata (Mart.) Becc. (Arecaceae), an iconic palm tree from the Brazilian semiarid region.

    Science.gov (United States)

    Simplicio, R R; Pereira, D G; Waldschmidt, A M

    2017-06-29

    The licuri palm Syagrus coronata plays a key role in the ecology and economy of Brazilian semiarid region. Nonetheless, genetic data about populations of this species are absent even though the intensive and uncontrolled exploitation since colonial periods has threatened the sustainability and viability of licuri populations. Therefore, we attempted to test the efficacy of transferability of microsatellite loci isolated from three palm tree species to S. coronata to analyze the population of this species throughout their range. A set of 19 heterologous microsatellite loci was tested in three native populations of S. coronata from the State of Bahia, northeastern Brazil, which amplified using distinct annealing temperatures (50°-60°C). Based on the 10 most polymorphic loci, the selected populations exhibited a mean number of alleles per locus of 9.8, and high genetic diversity values since the expected heterozygosity ranged from 0.573 to 0.754, while the observed heterozygosity varied from 0.785 to 1.000. In conclusion, the tested loci are transferrable and highly efficient to population studies in S. coronata, thus minimizing the lack of species-specific loci to the genetic monitoring of licuri populations.

  18. Seasonal and regional variations in the transfer of cesium radionuclides from soil to roe deer and plants in a prealpine forest

    International Nuclear Information System (INIS)

    Lindner, G.; Drissner, J.; Hund, M.; Zibold, G.; Zimmerer, R.; Herrmann, T.; Zech, W.

    1994-01-01

    The surveillance of roe deer contamination with cesium radionuclides in the rural prealpine area of Oberschwaben in south-west Germany since autumn 1986 revealed characteristic regional and seasonal patterns, which resulted from the transfer from soil to grazing plants of these animals. A periodic maximum in autumn was correlated with the mushroom season in forests. In the largest forest in this region, the transfer soil to roe deer was significantly higher in spruce than in mixed parts of this forest. In the soil in the spruce stand, the highest fraction of the total cesium radionuclide inventory was found in the O h horizon rich in organic matter. At an area in the spruce stand treated mainly with CaCO 3 in 1984, the transfer soil to plant was reduced by a factor of 4-5 compared to the neighbouring untreated area

  19. Air pollution and health implications of regional electricity transfer at generational centre and design of compensation mechanism

    Science.gov (United States)

    Relhan, Nemika

    India's electricity generation is primarily from coal. As a result of interconnection of grid and establishment of pithead power plants, there has been increased electricity transfer from one region to the other. This results in imbalance of pollution loads between the communities located in generation vis-a-vis consumption region. There may be some states, which are major power generation centres and hence are facing excessive environmental degradation. On the other hand, electricity importing regions are reaping the benefits without paying proper charges for it because present tariff structure does not include the full externalities in it. The present study investigates the distributional implications in terms of air pollution loads between the electricity generation and consumption regions at the state level. It identifies the major electricity importing and exporting states in India. Next, as a case study, it estimates the health damage as a result of air pollution from thermal power plants (TPPs) located in a critically polluted region that is one of the major generator and exporter of electricity. The methodology used to estimate the health damage is based on impact pathway approach. In this method, air pollution modelling has been performed in order to estimate the gridded Particulate Matter (PM) concentration at various receptor locations in the study domain. The air quality modeling exercise helps to quantify the air pollution concentration in each grid and also apportion the contribution of power plants to the total concentration. The health impacts as a result of PM have been estimated in terms of number of mortality and morbidity cases using Concentration Response Function (CRF's) available in the literature. Mortality has been converted into Years of Life Lost (YOLL) using life expectancy table and age wise death distribution. Morbidity has been estimated in terms of number of cases with respect to various health end points. To convert this health

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  1. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  2. Report on the PWR-radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Malone, D.J. [Consumers Power Co., Covert, MI (United States)

    1995-03-01

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.

  3. Three basic options for the management of PWR waste

    International Nuclear Information System (INIS)

    Malherbe, J.; Saulieu, E. de; Glibert, R.; Alamo Berna, S.; Cecille, L.; Geiser, H.; Kowa, S.; Thiels, G.

    1990-01-01

    Relying on the national practices of France, Germany and Belgium, three reference management routes for PWR wastes were drawn up and subsequently evaluated in terms of costs and radiological impact. It was thus demonstrated that safety regulations and technical redundancies, especially for off-gas treatment, liquid waste processing and dry solid waste treatment, play an important part in the cost associated with each route. The analysis of the different treatment options for mixed solid low level waste highlighted the low cost effectiveness of incineration as compared to compaction. Whatever the scenario investigated, the disposal costs of PWR wastes proved to be quite marginal in the overall cost. The radiological impact associated with each route was assessed through individual doses resulting from liquid and gaseous effluents. This theoretical exercise included some sensitivity studies performed on a selection of important parameters

  4. Thermal hydraulic simulations of the Angra 2 PWR

    Directory of Open Access Journals (Sweden)

    González-Mantecón Javier

    2015-01-01

    Full Text Available Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR. Simulations of the reactor behavior during steady state and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the developed RELAP5-3D model is capable of reproducing the thermal hydraulic behavior of the Angra 2 PWR and it can contribute to the process of the plant safety analysis.

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

    International Nuclear Information System (INIS)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de

    2015-01-01

    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)

  6. Genes encoding conserved hypothetical proteins localized in the conjugative transfer region of plasmid pRet42a from Rhizobium etli CFN42 participate in modulating transfer and affect conjugation from different donors.

    Directory of Open Access Journals (Sweden)

    Susana eBrom

    2015-01-01

    Full Text Available Among sequenced genomes, it is common to find a high proportion of genes encoding proteins that cannot be assigned a known function. In bacterial genomes, genes related to a similar function are often located in contiguous regions. The presence of genes encoding conserved hypothetical proteins (chp in such a region may suggest that they are related to that particular function. Plasmid pRet42a from Rhizobium etli CFN42 is a conjugative plasmid containing a segment of approximately 30 Kb encoding genes involved in conjugative transfer. In addition to genes responsible for Dtr (DNA transfer and replication, Mpf (Mating pair formation and regulation, it has two chp-encoding genes (RHE_PA00163 and RHE_PA00164 and a transcriptional regulator (RHE_PA00165. RHE_PA00163 encodes an uncharacterized protein conserved in bacteria that presents a COG4634 conserved domain, and RHE_PA00164 encodes an uncharacterized conserved protein with a DUF433 domain of unknown function. RHE_PA00165 presents a HTH_XRE domain, characteristic of DNA-binding proteins belonging to the xenobiotic response element family of transcriptional regulators. Interestingly, genes similar to these are also present in transfer regions of plasmids from other bacteria. To determine if these genes participate in conjugative transfer, we mutagenized them and analyzed their conjugative phenotype. A mutant in RHE_PA00163 showed a slight (10 times but reproducible increase in transfer frequency from Rhizobium donors, while mutants in RHE_PA00164 and RHE_PA00165 lost their ability to transfer the plasmid from some Agrobacterium donors. Our results indicate that the chp-encoding genes located among conjugation genes are indeed related to this function. However, the participation of RHE_PA00164 and RHE_PA00165 is only revealed under very specific circumstances, and is not perceived when the plasmid is transferred from the original host. RHE_PA00163 seems to be a fine-tuning modulator for conjugative

  7. Studi Operasi Resin Penukar Ion Dalam Sistem Purifikasi Air Primer Pwr

    OpenAIRE

    Biyantoro, Dwi; Basuki, Kris Tri; Subagiono, Subagiono

    2006-01-01

    STUDI OPERASI RESIN PENUKAR ION DALAM SISTEM PURIFIKASI AIR PRIMER PWR. Telah dilakukan studioperasi resin penukar ion dalam sistem purifikasi air primer PWR. Air pendingin reaktor yang pada awalnya sesuaidengan persyaratan setelah pengoperasian reaktor sering kualitasnya berubah, sehingga harus dimurnikan. Unsurunsurpengotor dalam air primer PWR diidentifikasi sebagai penyebab pengotor seperti korosi, pelepasan produk fisi(Cs137, Sr90, Co60,C14, Tc99), dan pelepasan kembali unsur oleh resin ...

  8. Uranium savings on a once through PWR fuel cycle

    International Nuclear Information System (INIS)

    Cupo, J.V.

    1980-01-01

    A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

  9. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    Matsumiya, Masayuki

    1996-01-01

    The main control room which is the information center of nuclear power plant has been continuously improved utilizing the state of the art ergonomics, a high performance computer, computer graphic technologies, etc. For the latest Japanese Pressurized Water Reactor (PWR) plant, the CRT monitoring system is applied as the major information source for facilitating operators' plant monitoring tasks. For an operating plant, enhancement of monitoring and logging functions has been made adopting a high performance computer

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

    OpenAIRE

    Huiun Ha; Sangwon Lee; Hangon Kim

    2017-01-01

    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 fo...

  11. Methodology of safety operating range determination for the PWR plant

    International Nuclear Information System (INIS)

    Kostadinov, V.; Mavko, B.

    1991-01-01

    The results of NPP Krsko core thermal power design limits investigation, which set bounds to the maximum allowable fuel temperature during normal operation and incidents of moderate frequency, are presented. In addition, allowable reactor coolant temperatures limited by the pressure of the steam generator safety valves opening are calculated. The range of a PWR plant safe operation imposed by the thermal overpower, the steam generator safety valves opening and DNBR safety limits is determined. (author)

  12. Fire experiences: principal lessons learned, application in PWR power plants

    International Nuclear Information System (INIS)

    Schoemacker, M.

    1984-01-01

    The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables [fr

  13. LOCA verification and data bank. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Varacalle, Jr., D. J.; Cox, N. D.; Atwood, C. L.; Madden, S. C.; Condie, K. G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique.

  14. PWR's countermeasures after Fukushima Daiichi Nuclear Power Plant accident

    International Nuclear Information System (INIS)

    Kato, Akihiko

    2012-01-01

    Countermeasures in case of similar events (loss of all AC power sources (SBO) and loss of ultimate heat sink (LUHS) as Fukushima Daiichi were investigated for further improvement of safety and reliability of PWR Plants. As for PWR in case of SBO and LUHS, steam driven auxiliary feedwater pump could be operable to supply feedwater to steam generators and stable state of reactor could be attained by natural circulation cooling of primary coolant Generated steam would be released to the air from main steam relief valve. Emergency safety countermeasures were taken to (1) improve water tightness by application of door and pipe penetration sealing to protect important equipment from flooding due to tsunami, (2) deploy mobile engine-operated pumps and (3) deploy emergency air-cooled generators. The government ordered 'stress tests' to quantify the effectiveness of safety measures for all Japan's reactors before they restart following any shutdown. Based on emergency safety countermeasures, plant operators assessed whether reactor and spent fuel pool could be stably cooled by external events (earthquake, tsunami and simultaneous effects) beyond the plant design basis. Further safety and reliability improvements of PWR plants had been considered or implemented for reinforcement of external power system, onsite power system with additional installation of permanent emergency air-cooled generators, enhancement of plant cooling function and update closure function of containment vessel in case of severe accident (T. Tanaka)

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

    International Nuclear Information System (INIS)

    Claro, Luiz H.

    2009-01-01

    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 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∞ 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)

  16. Evaluation of zinc addition to PWR primary coolant

    International Nuclear Information System (INIS)

    Pathania, R.; Yagnik, S.; Gold, R.E.; Dove, M.; Kolstad, E.

    1995-01-01

    Laboratory studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. Because of the potential benefits of zinc addition in reducing radiation fields and mitigating PWSCC of Alloy 600 a project was initiated to qualify zinc addition to a PWR. The objective of this work was to evaluate the effect of zinc addition on radiation fields, PWSCC of Alloy 600 and fuel cladding corrosion at the Farley-2 PWR. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor prior to zinc addition at Farley-2. This paper provides an overview of the scope of the zinc addition demonstration at Farley-2 and the fuel cladding corrosion tests at Halden. The zinc concentration in the Farley-2 coolant is approximately 40 ppb and that in Halden is 50 ppb. The paper presents initial results from these studies which are still in progress

  17. A scheme of better utilization of PWR spent fuels

    International Nuclear Information System (INIS)

    Chung, Bum Jin; Kang, Chang Soon

    1991-01-01

    The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

  18. 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.

  19. Electron transfer from plant phenolates to carotenoid radical cations. Antioxidant interaction entering the Marcus theory inverted region.

    Science.gov (United States)

    Cheng, Hong; Han, Rui-Min; Zhang, Jian-Ping; Skibsted, Leif H

    2014-01-29

    β-Carotene, lycopene, and zeaxanthin are maximally regenerated by plant phenolates from their radical cations formed during laser flash photolysis in 9:1 (v/v) chloroform/methanol for a driving force corresponding to the reorganization energy according to the Marcus theory. For β-carotene, the reorganization energy has values of 0.41 ± 0.04 and 0.40 ± 0.04 eV for the plant phenols in the presence of 1 and 2 equiv of base, respectively, at 23 °C. For a driving force lower than the reorganization energy, regeneration of the carotenoids is less efficient as is seen for m-hydroxybenzoic acid, vanillic acid, and p-coumaric acid. For a driving force above the maximum rate as determined to have kET = 6.3 × 10(9) L·mol(-1)·s(-1) for syringic acid and β-carotene, the reaction becomes gradually slower and regeneration less efficient as is seen for the more reducing caffeic acid, rutin, and quercetin corresponding to an inverted region for the rate of electron transfer. Lycopene and zeaxanthin show a similar behavior for the same series of plant phenols with slightly lower reorganization energy, in agreement with the lower reduction potential of their radical cations, while, for the ketocarotenoids astaxanthin and canthaxanthin, fast reactions with a solvent of radical cations inhibit regeneration from being detected. Intermediate reducing plant phenols accordingly yield maximal protection of carotenoids against photobleaching in foods and beverages.

  20. Fog inerting effects on hydrogen combustion in a PWR ice condenser contaminant

    International Nuclear Information System (INIS)

    Luangdilok, W.; Bennett, R.B.

    1995-01-01

    A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H 2 -air-steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser contaminant during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H 2 by volume). 18 refs., 3 tabs

  1. A digital simulation of a pressurizer in a PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sato, E.F.

    1980-11-01

    A model for pressurizer digital simulation of a PWR nuclear power plant during transients, considering all pressurizer control features, is presented. The pressurizer is divided into two regions separated by a water-vapor interface and non-equilibrium conditions are considered. The particular thermodynamic process followed during insurge and outsurges is determined at each instant of analysis without any previous assumption. The pressure behavior is defined by an explicit equation in any of four possible pressurizer thermodynamic conditions. Thermodynamic properties of steam and water are computed by ASME subroutines and the mathematical formulation presented in this study was programed in FORTRAN IV for a Burroughs-6700 digital computer system. This program was employed to simulate the Shippingport Atomic Power Station and Almirante Alvaro Alberto Nuclear Power Plant - Unit 1 pressurizers. The test results compared with experimental or vendor data show the validity of this analysis method. (Author) [pt

  2. Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems

    International Nuclear Information System (INIS)

    Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng

    2010-01-01

    A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)

  3. Transfer of the 3' non-translated region of grapevine chrome mosaic virus RNA-1 by recombination to tomato black ring virus RNA-2 in pseudorecombinant isolates.

    Science.gov (United States)

    Le Gall, O; Candresse, T; Dunez, J

    1995-05-01

    In grapevine chrome mosaic and tomato black ring viruses (GCMV and TBRV), as in many other nepoviruses, the 3' non-translated regions (3'NTR) are identical between the two genomic RNAs. We have investigated the structure of the 3'NTR of two recombinant isolates which contain GCMV RNA-1 and TBRV RNA-2. In these isolates, the 3'NTR of RNA-1 was transferred to RNA-2, thus restoring the 3' identity. The transfer occurred within three passages, and probably contributes to the spread of randomly appearing mutations from one genomic RNA to the other. The site of recombination is near the 3' end of the open reading frame.

  4. Hydro mechanical investigation on different PWR upper plenum core structures

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Ping'an; Yang Guanyue

    1997-01-01

    The development of Nuclear Industry relys on the safe and reliable operation of nuclear power station. Whether or not control rods moving upward and downward freedly and dropping rapidly in emergency case by order directly dominates the nuclear power regulation and emergency shut-down. So to clarify the factors which exert great influences on the drop of control rods is very important for making certain that PWR is operated safety and relialy. Among the factors, the hydraulic load on the control rods plays an important role during the operation of reactor. However because of complication in turbulent flow and concentration of the control rod guide bundles in the upper plenum, the flow field has not been thoroughly studied up to now. In order to understand the flow field in upper plenum fully a 1/4 scale transparent model of the upper plenum of a active 300 MWe PWR is designed and installed in line with similarity theory. The velocity distributions (including horizontal and axial velocity) in the upper plenum are obtained by using N-J type Dynamic Resistance Strain Foil Velocimetry (N-J type DRSFV) and Laser Doppler Velocimetry (LDV). For the sake of alleviating the hydraulic load on the control rods and making certain that the control rods and making certain that the control rods are moving upward and downward freely and drop rapidly in emergency case by order, the core structure in the upper plenum of the active 300 MWe PWR is improved as in the following 2 cases: 1 Some protective sleeves are added to the control rod guide bundles near the upper plenum outlet nozzles (symmetric 4 bundles: 02-26, 03-25, 11-29, 12-28). The rest of the core structure is same as that of the core structure in the active 300 MWe PWR. 2. The active upper plenum core structure with 37 control rod guide bundles is replaced by the core structure with 33 protective-sleeved control rod guide bundles. The results of the simulated experiments with the 2 cases are compared with that of the

  5. 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

  6. Secondary flow and heat transfer coefficient distributions in the developing flow region of ribbed turbine blade cooling passages

    Science.gov (United States)

    Forsyth, Peter; McGilvray, Matthew; Gillespie, David R. H.

    2017-01-01

    This paper reports an experimental and numerical study of the development and coupling of aerodynamic flows and heat transfer within a model ribbed internal cooling passage to provide insight into the development of secondary flows. Static instrumentation was installed at the end of a long smooth passage and used to measure local flow features in a series of experiments where ribs were incrementally added upstream. This improves test turnaround time and allows higher-resolution heat transfer coefficient distributions to be captured, using a hybrid transient liquid crystal technique. A composite heat transfer coefficient distribution for a 12-rib-pitch passage is reported: notably the behaviour is dominated by the development of the secondary flow in the passage throughout. Both the aerodynamic and heat transfer test data were compared to numerical simulations developed using a commercial computational fluid dynamics solver. By conducting a number of simulations it was possible to interrogate the validity of the underlying assumptions of the experimental strategy; their validity is discussed. The results capture the developing size and strength of the vortical structures in secondary flow. The local flow field was shown to be strongly coupled to the enhancement of heat transfer coefficient. Comparison of the experimental and numerical data generally shows excellent agreement in the level of heat transfer coefficient predicted, though the numerical simulations fail to capture some local enhancement on both the ribbed and smooth surfaces. Where this was the case, the coupled flow and heat transfer measurements were able to identify missing velocity field characteristics.

  7. Flexibility control and simulation with multi-model and LQG/LTR design for PWR core load following operation

    International Nuclear Information System (INIS)

    Li, Gang; Zhao, Fuyu

    2013-01-01

    Highlights: ► The nonlinear model and linear multi-model of a PWR core are developed. ► The LQG/LTR robust control is used to design local controllers of the core. ► LTR principles are analyzed and proved theoretically. ► Flexibility control is proposed to design flexibility controllers for the core. ► The nonlinear core load following control system is effective. - Abstract: The objective of this investigation is to design a nonlinear Pressurized Water Reactor (PWR) core load following control system. On the basis of modeling a nonlinear PWR core, linearized models of the core at five power levels are chosen as local models of the core to substitute the nonlinear core model in the global range of power level. The Linear Quadratic Gaussian with Loop Transfer Recovery (LQG/LTR) robust optimal control is used to contrive a controller with the robustness of a core local model as a local controller of the nonlinear core. Meanwhile, LTR principles are analyzed and proved theoretically by adopting the matrix inversion lemma. Based on the local controllers, the principle of flexibility control is presented to design a flexibility controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. Finally, the core load following control system is simulated and the simulation results show that the control system is effective

  8. Development, income transfer strategies, and the nutritional transition in Brazilian children from a rural and remote region.

    Science.gov (United States)

    Freitas, D A; Sousa, Á A; Jones, K M

    2014-01-01

    Global development processes have been associated with the nutritional transition, where undernutrition is replaced by overnutrition. Income transfer policies in Brazil have targeted hunger, but may not address the need for balanced nutrition. Data was collected from government databanks that document the nutritional status of Brazilians applying for social services. This data was analyzed for descriptive statistics. Development and income transfer processes appear to be associated with an increase in overweight children between the years 2008 and 2012. Income transfer programs need to incorporate educational programs that address the need to budget for balanced nutrition.

  9. Development of an ultrasonic inspection technique for the Sizewell B PWR cast austenitic reactor coolant pump bowl

    International Nuclear Information System (INIS)

    Gilroy, K.S.

    1988-01-01

    The CEGB has recently received approval to build a PWR at Sizewell in Suffolk. The CEGB's Non-Destructive Testing Applications Centre is developing ultrasonic inspection techniques for those components of the primary circuit whose failure is claimed to be incredible. For the welds and castings that fall into this category, the ultrasonic inspections will be validated using test-blocks of representative macrostructure and geometry, and containing realistic defects. This paper describes the research leading to the development of an ultrasonic inspection technique for the near-surface regions of the cast austenitic reactor coolant pump bowls. (author)

  10. Search for proton emission in {sup 54}Ni and multi-nucleon transfer reactions in the actinide region

    Energy Technology Data Exchange (ETDEWEB)

    Geibel, Kerstin

    2012-06-15

    The first part of the thesis presents the investigation of fusion-evaporation reactions in order to verify one-proton emission from the isomeric 10{sup +} state in the proton rich nucleus {sup 54}Ni. Between the years 2006 and 2009 a series of experimental studies were performed at the Tandem accelerator in the Institut fuer Kernphysik (IKP), University of Cologne. These experiments used fusion-evaporation reactions to populate {sup 54}Ni via the two-neutron-evaporation channel of the compound nucleus {sup 56}Ni. The cross section for the population of the ground state of {sup 54}Ni was predicted to be in orders of microbarn. This required special care with respect to the sensitivity of the experimental setup, which consisted of a double-sided silicon-strip detector (DSSSD), a neutron-detector array and HPGe detectors. In two experiments the excitation functions of the reactions ({sup 32}S+{sup 24}Mg) and ({sup 28}Si+{sup 28}Si) were determined to find the optimal experimental conditions for the population of {sup 54}Ni. A final experiment employed a {sup 28}Si beam at an energy of 70 MeV, impinging on a {sup 28}Si target. With a complex analysis it is possible to obtain a background-free energy spectrum of the DSSSD. An upper cross section limit for the population of the 10{sup +} state in {sup 54}Ni is established at σ({sup 54}Ni(10{sup +})) ≤ (13.9 ± 7.8) nbarn. In the second part of the thesis the population of actinide nuclei by multi-nucleon transfer reactions is investigated. Two experiments, performed in 2007 and 2008 at the CLARA-PRISMA setup at the Laboratori Nazionali di Legnaro, are analyzed with respect to the target-like reaction products. In both experiments {sup 238}U was used as target. A {sup 70}Zn beam with 460 MeV and a {sup 136}Xe beam with 926 MeV, respectively, impinged on the target, inducing transfer reactions. Kinematic correlations between the reaction partners are used to obtain information on the unobserved target-like reaction

  11. Minimal and contributing sequence determinants of the cis-acting locus of transfer (clt) of streptomycete plasmid pIJ101 occur within an intrinsically curved plasmid region.

    Science.gov (United States)

    Ducote, M J; Prakash, S; Pettis, G S

    2000-12-01

    Efficient interbacterial transfer of streptomycete plasmid pIJ101 requires the pIJ101 tra gene, as well as a cis-acting plasmid function known as clt. Here we show that the minimal pIJ101 clt locus consists of a sequence no greater than 54 bp in size that includes essential inverted-repeat and direct-repeat sequences and is located in close proximity to the 3' end of the korB regulatory gene. Evidence that sequences extending beyond the minimal locus and into the korB open reading frame influence clt transfer function and demonstration that clt-korB sequences are intrinsically curved raise the possibility that higher-order structuring of DNA and protein within this plasmid region may be an inherent feature of efficient pIJ101 transfer.

  12. SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

    International Nuclear Information System (INIS)

    Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

    2014-01-01

    The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

  13. Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

    International Nuclear Information System (INIS)

    2003-11-01

    In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

  14. Fuel assembly for PWR type reactor

    International Nuclear Information System (INIS)

    Yokoyama, Takashi.

    1991-01-01

    In a fuel assembly in a reactor-loaded state, pellets to be loaded in a region higher than a predetermined height are made hollow. That is, the volume of the gap in the hollow portion of the pellet comprising fissible materials to be filled at a position higher by 1/2 to 1/3 height from the upper region (downstream of coolant flow) of at least a portion of fuel rods in a fuel assembly is reduced stepwise than that in the lower region (upstream of coolant flow). Alternatively, the volume of the gap is gradually reduced from the lower portion to the upper portion. The diameter of the hollow hole portion is thus varied and the volume of the gap of the fissible materials per pellet is controlled, to reduce the amount of the fissile materials per unit volume. With such a constitution, the power of fission energy in the reactor core upper portion can be lowered. Accordingly, the corrosion of a cladding tube upon high burnup degree can be suppressed, thereby enabling to ensure integrity. (T.M.)

  15. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  16. Soil to leaf transfer factor for the radionuclides {sup 226}Ra, {sup 40}K, {sup 137}Cs and {sup 90}Sr at Kaiga region, India

    Energy Technology Data Exchange (ETDEWEB)

    James, Joshy P., E-mail: jpjames@npcil.co.in [Environmental Survey Laboratory, Kaiga Generating Station, Health Physics Division, BARC, Kaiga, Uttar Kannada District, Karwar, Karnataka 581400 (India); Dileep, B.N. [Environmental Survey Laboratory, Kaiga Generating Station, Health Physics Division, BARC, Kaiga, Uttar Kannada District, Karwar, Karnataka 581400 (India); Ravi, P.M. [Health Physics Division, Bhabha Atomic Research Centre (BARC), Mumbai 400 085 (India); Joshi, R.M.; Ajith, T.L. [Environmental Survey Laboratory, Kaiga Generating Station, Health Physics Division, BARC, Kaiga, Uttar Kannada District, Karwar, Karnataka 581400 (India); Hegde, A.G.; Sarkar, P.K. [Health Physics Division, Bhabha Atomic Research Centre (BARC), Mumbai 400 085 (India)

    2011-12-15

    Transfer factors are the most important parameters required for mathematical modeling used for environmental impact assessment of radioactive contamination in the environment. In this paper soil to leaf transfer factor for the radionuclides {sup 40}K, {sup 226}Ra, {sup 137}Cs and {sup 90}Sr is estimated for Kaiga region in Karnataka state, India. Among the plants in which study is carried out, {sup 226}Ra, {sup 40}K, {sup 137}Cs and {sup 90}Sr activity in leaves of herbaceous plants is higher than that of tree leaves. Soil to leaf transfer factor for {sup 226}Ra, {sup 40}K, {sup 137}Cs and {sup 90}Sr was found to be in the range of 0.03-0.65, 0.32-8.04, 0.05-3.03 and 0.42-2.67 respectively. - Highlights: > Study region is Kaiga, Karnataka, India. > {sup 226}Ra, {sup 40}K, {sup 137}Cs and {sup 90}Sr activity in soil and leaf samples are reported. > {sup 226}Ra, {sup 40}K, {sup 137}Cs activity is higher in herbaceous plant leaves, than tree leaves. > In most of the plants transfer factor varies in the order K > Sr > Cs > Ra.

  17. Soil to leaf transfer factor for the radionuclides 226Ra, 40K, 137Cs and 90Sr at Kaiga region, India

    International Nuclear Information System (INIS)

    James, Joshy P.; Dileep, B.N.; Ravi, P.M.; Joshi, R.M.; Ajith, T.L.; Hegde, A.G.; Sarkar, P.K.

    2011-01-01

    Transfer factors are the most important parameters required for mathematical modeling used for environmental impact assessment of radioactive contamination in the environment. In this paper soil to leaf transfer factor for the radionuclides 40 K, 226 Ra, 137 Cs and 90 Sr is estimated for Kaiga region in Karnataka state, India. Among the plants in which study is carried out, 226 Ra, 40 K, 137 Cs and 90 Sr activity in leaves of herbaceous plants is higher than that of tree leaves. Soil to leaf transfer factor for 226 Ra, 40 K, 137 Cs and 90 Sr was found to be in the range of 0.03-0.65, 0.32-8.04, 0.05-3.03 and 0.42-2.67 respectively. - Highlights: → Study region is Kaiga, Karnataka, India. → 226 Ra, 40 K, 137 Cs and 90 Sr activity in soil and leaf samples are reported. → 226 Ra, 40 K, 137 Cs activity is higher in herbaceous plant leaves, than tree leaves. → In most of the plants transfer factor varies in the order K > Sr > Cs > Ra.

  18. EMERALD-NORMAL, Routine Radiation Release and Dose for PWR Design Analysis and Operation Analysis

    International Nuclear Information System (INIS)

    Gillespie, S.G.; Brunot, W.K.

    1976-01-01

    1 - Description of problem or function: EMERALD-NORMAL is designed for the calculation of radiation releases and exposures resulting from normal operation of a large pressurized water reactor. The approach used is similar to an analog simulation of a real system. Each component or volume in the plant which contains a radioactive material is represented by a subroutine which keeps track of the production, transfer, decay, and absorption of radioactivity in that volume. During the course of the analysis, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place in the plant. Some of this activity is then released to the atmosphere and to the discharge canal. The rates of transfer, leakage, production, cleanup, decay, and release are read as input to the program. Subroutines are also included which calculate the off-site radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for the forty isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD-NORMAL program can be used for most calculations involving the production and release of radioactive material. These include design, operation, and licensing studies. 2 - Method of solution: Explicit solutions of first-order linear differential equations are included. In addition, a subroutine is provided which solves a set of simultaneous linear algebraic equations. 3 - Restrictions on the complexity of the problem: Many parameters and systems included in the program, particularly the radiation waste-treatment system, are unique to the PG and E Diablo Canyon PWR plant. Maxima of: 50 isotopes, 9 distances, 16 angular sectors, 1 operating period, 1 reactor power level

  19. Safety aspects of the using Gd as burnable poison in PWR's

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.

    1978-01-01

    The experience of BELGONUCLEAIRE in using Gd in LWR's has indicated the safety related advantages of this burnable poison. The successfully operation of the BR3 PWR power plant with 5% of Gd rods is presented and extrapolated to large PWR's. (authro)

  20. PWR upper/lower internals shield

    Energy Technology Data Exchange (ETDEWEB)

    Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  1. Transfers of embodied PM2.5emissions from and to the North China region based on a multiregional input-output model.

    Science.gov (United States)

    Yang, Xue; Zhang, Wenzhong; Fan, Jie; Yu, Jianhui; Zhao, Hongyan

    2018-04-01

    Atmospheric PM 2.5 pollution has become a global issue, and is increasingly being associated with social unrest. As a resource reliant local economy and heavy industry cluster, the North China region has become China's greatest emitter, and the source of much pollution spillover to outside regions. To address this issue, the current study investigates the transfers of embodied PM 2.5 emissions to and from the North China region (which is taken to include Hebei, Henan, Shandong, and Shanxi, and is referred to here as HHSS). The study uses a top-down pollutant emission inventory and environmentally extended multi-regional input-output (EE-MRIO) model. The results indicate that the HHSS area exported a total of 660 Gg of embodied PM 2.5 to other domestic provinces, mainly producing outflows to China's central coastal area (Jiangsu, Zhejiang, and Shanghai) and the Beijing-Tianjin region. HHSS also imported 224 Gg of embodied PM 2.5 from other domestic regions, primarily from Inner Mongolia and the northeast. Furthermore, the transfer of embodied emissions often occurred between geographically adjacent areas to save costs; Beijing and Tianjin mainly transferred embodied pollution to Hebei and Shanxi, whilst Jiangsu, Shanghai, and Zhejiang tended to import embodied air pollutants from Shandong and Henan. At the sectoral level, the melting and pressing of metals, the production of non-metallic products, and electric and heat power production were the three dominant economic sectors for PM 2.5 emissions, together accounting for 81% of total discharges. Capital formation played a key role in outflows (75%) in all sectors. Moreover, the virtual pollutant emissions exported to foreign countries also significantly affected HHSS' discharges significantly, making up 340 Gg. Allocating responsibility for some proportion of HHSS' emissions to the Beijing-Tianjin area and the central coastal provinces may be an effective approach for mitigating releases in HHSS. Copyright

  2. Development of CHF correlation “MG-NV” for low pressure and low velocity conditions applied to PWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yumura, T.; Yodo, T.; Makino, Y.; Suemura, T. [Mitsubishi Heavy Industries, LTD., Kobe, Hyogo (Japan)

    2011-07-01

    The Critical Heat Flux (CHF) is one of the important parameters in the safety analysis of Pressurized Water Reactor (PWR). If the CHF is reached, an abrupt drop occurs in the heat transfer between the fuel rod cladding and the reactor coolant, which may induce a large temperature excursion of fuel cladding and a subsequent fuel failure. Therefore, accurate prediction of CHF is required in order to assure a sufficient safety margin in the PWR core. Mitsubishi Heavy Industries, ltd (MHI) is developing a new series of CHF correlations which covers various fuel designs and wide range of fluid conditions with sufficient reliability. In this paper, a new CHF correlation, MG-NV (Mitsubishi Generalized correlation for Non-Vane grid spacers) is presented. This correlation is one of the basic components of the new correlation series and was developed to cover low pressure and low velocity conditions where the rod bundle CHF data are limited. The CHF correlation was developed based on open CHF database and provides conservative but more reliable rod bundle CHF predictions compared with the conventional CHF correlations used in safety analyses at low pressure condition, such as Main Steam Line Break event. (author)

  3. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  4. PWR fuel pin diameter optimisation studies and economic analyses for uranium nitride fuel - 5048

    International Nuclear Information System (INIS)

    Thomas, G.M.; Grove, C.

    2015-01-01

    Alternative advanced fuels are currently being investigated by the nuclear industry. For example, research is underway into the possibility of replacing industry standard UO 2 fuel with Accident Tolerant Fuels (ATF) such as uranium nitride (UN). The higher density of UN compared with UO 2 results in a reduction in neutron moderation due to the lower hydrogen to heavy metal ratio (H/HM) for a given fuel assembly geometry in water. This suggests a different optimum UN fuel pin diameter in order to maximise lifetime average reactivity. If a smaller UN pellet/cladding diameter is adopted then the H/HM ratio is increased, leading to an increase in reactivity at lower burnup (followed by a reduction at higher burnup due to reduced Pu production). Preliminary studies have also indicated that a reduction of the UN pellet diameter with respect to standard UO 2 fuel could be beneficial to economic performance. This paper describes an approach used to determine the optimum fuel pin diameter for UN fuel in an AP1000 PWR using Studsvik CASMO4/SIMULATE3 neutronics codes. The objective is to maximise the fuel's lifetime average reactivity while staying within typical PWR nuclear design safety limits. The calculations demonstrate that the pin diameter should be decreased to optimise the fuel reactivity. However, if the pin diameter is decreased too much a highly undesirable positive moderator temperature coefficient can result. Economics calculations show that if UN fuel is used there is a potential economic benefit - in the region of 3 million dollar per 18-month reload if generic openly available cost data is used

  5. Study of power peak migration due to insertion of control bars in a PWR reactor

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Costa, Danilo Leite; Borges, Diogo da Silva; Lava, Deise Diana; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to present a study on the power distribution behavior in a PWR reactor, considering the intensity and the migration of power peaks as is the insertion of control rods in the core banks. For this, the study of the diffusion of neutrons in the reactor was adopted by computer simulation that uses the finite difference method for numerically solving the neutron diffusion equation to two energy groups in steady state and in symmetry of a fourth quarter core. We decided to add the EPRI-9R 3D benchmark thermal-hydraulic parameters of a typical power PWR. With a new configuration for the reactor, the positions of the control rods banks were also modified. Due to the new positioning of these banks in the reactor, there was intense power gradients, favoring the occurrence of critical situations and logically unconventional for operation of a nuclear reactor. However, these facts have led interesting times for the study on the power distribution behavior in the reactor, showing axial migration of power peaks and mainly the effect of the geometry of the core on the latter. Based on the distribution of power was evident the increase of the power in elements located in the central region of the reactor core and, concomitantly, the reduction in elements of its periphery. Of course, the behavior exhibited by the simulated reactor is not in agreement with that expected in an actual reactor, where the insertion of control rods banks should lead to reduced power throughout the core as evenly as possible, avoiding sharp power peaks, standardizing the burning fuel, controlling reactivity deviations and acting in reactor shutdown

  6. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  7. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  8. Thermal hydraulic design of hydride fueled PWR cores

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Romano, A.

    2004-01-01

    The neutronic characteristics of hydride fuels permit increased fuel to coolant volume ratios in the core. A parametric study was developed to determine the optimum combination of lattice pitch, rod diameter, and channel shape - further referred to as geometry - for minimizing the total cost of operating existing PWRs loaded with UZrH 1.6 fuel. Results of the thermal hydraulic and fuel performance studies are presented here, and will be integrated into an economic model in the next stage of the research. The thermal hydraulic analysis was used to determine the maximum power that can be achieved by a given geometry, subject to four constraints - MDNBR, pressure drop, fuel temperature, and coolant flow velocity. The fuel performance analysis was used to determine the maximum burnup that can be achieved by a given geometry, subject to three additional constraints - fuel internal pressure and fission gas release, clad oxidation, and clad strain. This methodology was successfully validated by comparison of the predicted power and burnup of the current PWR geometry, with the actual power and burnup of an existing PWR. Assuming a 60 psia pressure drop can be sustained through the fuel bundle, we concluded the following for square channels: the peak achievable power is 5556 MWt for a rod diameter of 6.5 mm and a P/D ratio of 1.43, and the highest power that can be achieved using the existing 12.6 mm pitch and 10.2 mm fuel rods is 4586 MWt. These power levels are significantly higher than the 3800 MWt of the reference PWR. (author)

  9. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  10. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  11. Multi-region fuzzy logic controller with local PID controllers for U-tube steam generator in nuclear power plant

    Directory of Open Access Journals (Sweden)

    Puchalski Bartosz

    2015-12-01

    Full Text Available In the paper, analysis of multi-region fuzzy logic controller with local PID controllers for steam generator of pressurized water reactor (PWR working in wide range of thermal power changes is presented. The U-tube steam generator has a nonlinear dynamics depending on thermal power transferred from coolant of the primary loop of the PWR plant. Control of water level in the steam generator conducted by a traditional PID controller which is designed for nominal power level of the nuclear reactor operates insufficiently well in wide range of operational conditions, especially at the low thermal power level. Thus the steam generator is often controlled manually by operators. Incorrect water level in the steam generator may lead to accidental shutdown of the nuclear reactor and consequently financial losses. In the paper a comparison of proposed multi region fuzzy logic controller and traditional PID controllers designed only for nominal condition is presented. The gains of the local PID controllers have been derived by solving appropriate optimization tasks with the cost function in a form of integrated squared error (ISE criterion. In both cases, a model of steam generator which is readily available in literature was used for control algorithms synthesis purposes. The proposed multi-region fuzzy logic controller and traditional PID controller were subjected to broad-based simulation tests in rapid prototyping software - Matlab/Simulink. These tests proved the advantage of multi-region fuzzy logic controller with local PID controllers over its traditional counterpart.

  12. 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)

  13. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    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

  14. Natural-circulation-cooling characteristics during PWR accident simulations

    International Nuclear Information System (INIS)

    Adams, J.P.; McCreery, G.E.; Berta, V.T.

    1983-01-01

    A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

  15. Conversion ratio in epithermal PWR, in thorium and uranium cycle

    International Nuclear Information System (INIS)

    Barroso, D.E.G.

    1982-01-01

    Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.) [pt

  16. Sizewell B - analysis of British application of US PWR technology

    International Nuclear Information System (INIS)

    1983-05-01

    This report provides information on the staff's evaluation of major design differences and issues developed by the British in their application (Sizewell B) of US PWR technology. One design change, the addition of steam-driven charging pumps, was assessed to have a relatively high value compared to the other changes. However, the assessment is based on a number of assumptions for which inadequate data exist to make an unqualified judgment. Other changes to the US design (as typified by the SNUPPS design) were found to have relatively low or moderate safety benefits for US application

  17. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  18. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  19. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  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. Structure-dynamic model verification calculation of PWR 5 tests

    International Nuclear Information System (INIS)

    Engel, R.

    1980-02-01

    Within reactor safety research project RS 16 B of the German Federal Ministry of Research and Technology (BMFT), blowdown experiments are conducted at Battelle Institut e.V. Frankfurt/Main using a model reactor pressure vessel with a height of 11,2 m and internals corresponding to those in a PWR. In the present report the dynamic loading on the pressure vessel internals (upper perforated plate and barrel suspension) during the DWR 5 experiment are calculated by means of a vertical and horizontal dynamic model using the CESHOCK code. The equations of motion are resolved by direct integration. (orig./RW) [de

  2. Improvement in PWR flexibility the french program 1975-1995

    International Nuclear Information System (INIS)

    Gautier, A.; Miossec, C.

    1985-12-01

    Between 1975 and 1985, a substantial effort was launched in France to greatly improve PWR's flexibility, resulting in the current situation where both frequency control and load follow are now routinely performed on most plants in operation. Based on rapidly accumulating operational experience and on all expertise acquired in the past decade, a second-generation core control strategy is now being finalized for application on all future 1400 MW plants (with commercial operation scheduled in 1992 for first unit). This 20-year program is discussed

  3. New genetic algorithms (GA) to optimize PWR reactors

    International Nuclear Information System (INIS)

    Alim, Fatih; Ivanov, Kostadin; Levine, Samuel H.

    2008-01-01

    The objective of this study was to develop a unique scientific methodology as well as a practical tool for designing the loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and corresponding BP placement design that will achieve maximum cycle length while satisfying the safety constraints. To solve this optimization problem, a core reload optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed. This code is applicable for all types of PWR cores having different geometries and designs with an unlimited number of FA types in the inventory. GARCO has three modes: the user can optimize the core configuration (LP pattern) with or without BPs in the first mode; the second mode is the optimization of BP placement in the core and the last mode is the user can optimize LP and BP placements simultaneously in mode 3. In this study, the first mode finds the optimal LPs using the Haling Power Depletion Method (HPD) for placing BPs in the core. The second mode, which depletes the core accurately, places BPs in the selected optimum LP pattern. This methodology is applied only to the TMI-1 PWR. However, the improved Mode 1 GA option was applied to both the VVER-1000 and the TMI-1 to demonstrate and verify the advantages of the new enhancements in optimizing the LP pattern only. The 'Moby-Dick' code is used as reactor physics code for VVER-1000 analysis in this research. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1. The libraries of the BP designs used in SIMULATE-3 in this study were produced by Yilmaz (2005) [Yilmaz, S., 2005. Multilevel optimization of burnable poison utilization for advanced PWR fuel management. Ph.D. Thesis in

  4. Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors

    International Nuclear Information System (INIS)

    Boulot, F.

    1993-01-01

    Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs

  5. Electropolishing process development for PWR steam generator channel heads

    International Nuclear Information System (INIS)

    Asay, R.H.; Graves, P.; Guastaferro, C.T.; Spalaris, C.N.

    1991-04-01

    A broad range of process parameters was established to smoothen the surface of 309 L weld clad overlay, prototypic of surfaces common is channel heads of replacement PWR [pressurized water reactor] steam generators. Mechanical and electropolishing steps were studied to explore process boundaries, which result in acceptable degree of surface smoothness, without compromising metallurgical properties. Recommended processes and acceptance criteria established in this work, can be applied to electropolish steam generator channel heads. Smooth surfaces are less likely to retain radioactive species, and potentially develop lower radiation fields when these components are placed into service. 7 refs., 11 figs., 12 tabs

  6. Radiation protection optimization in the PWR type reactor dismantling

    International Nuclear Information System (INIS)

    Hilmoine, R.

    1998-01-01

    The studies made at the international level for the PWR type reactors, give dosimetric evaluations about 10 to 15 h.Sv for an immediate dismantling and around three to four times lower for a delayed dismantling according to the storage time. The technical hypothesis, the ambient dosimetry, the time of occupational exposure and the radioactive wastes management are not clearly specified so, Electricite de France has undertaken a more exhaustive study that takes into account, the radiation protection dimension in its universality from a complete radiological characterization in a standard installation. (N.C.)

  7. Chemical cleaning of PWR steam generators: application at Nogent 1

    International Nuclear Information System (INIS)

    Fiquet, J.M.; Veysset, J.P.; Esteban, L.; Saurin, P.

    1990-01-01

    EDF has developed and patented a chemical cleaning process for PWR steam generators, based on the use of a mixture of organic acids in order to: - dissolve iron oxides and copper with a single solution; - clean dented crevices. Qualification tests have permitted to demonstrate effectiveness of the solution and its inocuousness related to steam generator materials. The process, the license of which belongs to SOMAFER R.A. and FRAMATOME, has been implemented in France at Nogent. The goal was to dissolve iron oxides allowing metallic particles, aggregated on the tubesheet, to be released and mechanically removed. The effectiveness was satisfactory and this treatment is to be extended to other units [fr

  8. Life management plants at nuclear power plants PWR

    International Nuclear Information System (INIS)

    Esteban, G.

    2014-01-01

    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)

  9. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  10. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    Esteves, A.M.; Silva, A.T. e.

    1992-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  11. The N4 plant: culmination of French PWR experience

    International Nuclear Information System (INIS)

    Bellet, J.; Houyez, A.; Journet, J.; Pierrard, J.H.

    1985-01-01

    The model N4 series of 1400MWe class PWR plants has been developed in France from a unique base of technical and operating experience. It meets the French government's requirement for a reactor free of constraints due to licensing agreements with overseas companies, with enhanced safety features and incorporating the lessons of Three Mile Island. In particular, improvements have been made to the reactor vessel, the steam generators, the primary pumps and control systems. The units are capable of daily load following and extended operation between refuelling. The N4 plant includes a new design of turbine-generator. (author)

  12. Substitution of cobalt alloying in PWR primary circuit gate valves

    International Nuclear Information System (INIS)

    Cachon, L.; Sudreau, F.; Brunel, L.

    1995-01-01

    The object of this study is qualify cobalt-free alternative alloys for valve applications. This paper focus on tribological characterization of numerous coatings is done by using the first one, of a classical type. Then tests are performed with the second one which simulates solicitations supported by gate valves in primary circuit of PWR. 35% Ni-Cr - 65% Cr 3 C 2 coating, deposited by detonation gun technology, gives us hope to find a substitute of Stelite 6. (author). 5 refs., 16 figs., 2 tabs

  13. Vertical Drop Of 21-PWR Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

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

    2001-01-01

    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

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

    Directory of Open Access Journals (Sweden)

    Endiah Puji Hastuti

    2015-04-01

    yang sangat lama dan membutuhkan memori yang besar. Kata kunci: aliran turbulen, kanal PWR, CFD, tunak, transien   Coolant flow turbulence on heat transfer process serves to enhance the heat transfer coefficient, likewise flow in the fuel sub channel. Computational fluid dynamic program, FLUENT is a computational program based on finite element, that is able to predict and analyze the dynamics of fluid flow phenomena, accurately. CFD calculation program is selected in this study because of its accurately and it also can provide good visualization. Purpose of this research was to understand the characteristics of heat transfer, mass and momentum of the fuel rod to the coolant visually on: the temperature field, pressure field, and the kinetic energy field, as a function of the flow dynamics within fuel channel, on steady state and transient condition. Analysis of flow dynamics in the fuel channel base on CFD was done by using the PWR sample data with reactor power of 1000 MWe on 17x17 array of fuel. To examine the sensitivity of the flow equation in accordance with the model of turbulent flow on fuel channel, the turbulence equation model of k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, and Reynold stress model (RSM for steady state was used, while for transient turbulence model DES and LES are applied. In the sensitivity analysis of turbulent flow, hexahedral mesh model of three cell geometry each are 0.5 mm, 0.2 mm and 0.15 mm, was selected. The analysis shows that there are similar results of turbulen model Ƙ-ε and Ƙ-ω standard, on steady state analysis. Comparing with Dittus Boelter criteria for Nusselt number, the Reynolds stress model (RSM is recommended. Sensitivity analysis of mesh geometry between cell size 0.5 mm, 0.2 mm and 0.15 mm, indicating that the cell size of 0.5 mm was sufficient. Developed flow already reached on DES and LES model, however only for short time (3 seconds for transient condition. LES model need very long computation time and big memory

  15. Development of the business area construction and energy of EnergieRegion Nuernberg. Transfer from project management to a regional network

    International Nuclear Information System (INIS)

    Seiverth, A.

    2006-01-01

    The association EnergieRegion Nuernberg is a regional authority network, which is employed with the promotion of sustainable handling of the factor energy in the region Nuernberg and with the proliferation of this region as internationally recognized location for energy engineering, energy industry and energy science. The intention is to use the important industrial, service-oriented and scientific potential optimally. For this reason a functional co-ordination and communication platform had to be created for the cross-linking of the appropriate participants from economics, research and public administration. Therefore, the author of the contribution under consideration accompanies the development process of the business field construction and energy of this association in the background of the current trends in the construction and energy sector in the region Nuernberg. Under this aspect, the author reports on the following aspects: (a) Success factors of the project management in a regional network; (b) Operationalisation of the success of the project by means of a model; (c) Analysis of the different aspects of energetic measures; (d) Determination of chances and risks of the range building and energy in the region Nuernberg; (e) Comparison of the success of the model projects with the model for the determination of project success; (f) Determination of strengths and weaknesses of the project management in the business field construction and energy of the energy region Nuernberg

  16. International technology transfer

    International Nuclear Information System (INIS)

    Kwon, Won Gi

    1991-11-01

    This book introduces technology progress and economic growth, theoretical consideration of technology transfer, policy and mechanism on technology transfer of a developed country and a developing country, reality of international technology transfer technology transfer and industrial structure in Asia and the pacific region, technology transfer in Russia, China and Eastern Europe, cooperation of science and technology for development of Northeast Asia and strategy of technology transfer of Korea.

  17. Post-CHF low-void heat transfer of water: measurements in the complete transition boiling region at atmospheric pressure

    International Nuclear Information System (INIS)

    Johannsen, K.; Meinen, W.

    1984-01-01

    An experimental investigation of low-void heat transfer of water has been performed in the range of CHF and the minimum stable film boiling temperature. The heat transfer system used consists of a vertically mounted copper tube of 1 cm I.D. and 5 cm length with surface-temperature controlled, indirect Joule heating. Results are presented for upflowing water at inverted annular flow conditions in the inlet subcooling range of 2.5 - 40 0 C and mass flux range of 137-600 kg/m 2 s in terms of boiling curves and heat transfer coefficients versus wall temperature. Heat transfer in the stationary rewetting front, which occurs within the test section during operation in the transition boiling mode, is also dealt with. At high mass flux, occurrence of an inverse rewetting front has been observed. It is also noted that, at fixed location, minimum heat flux observed is usually not associated with the minimum stable film boiling temperature

  18. Signal-transfer Modeling for Regional Assessment of Forest Responses to Environmental Changes in the Southeastern United States

    Science.gov (United States)

    Robert J. Luxmoore; William W. Hargrove; M. Lynn Tharp; Wilfred M. Post; Michael W. Berry; Karen S. Minser; Wendell P. Cropper; Dale W. Johnson; Boris Zeide; Ralph L. Amateis; Harold E. Burkhart; V. Clark Baldwin; Kelly D. Peterson

    2000-01-01

    Stochastic transfer of information in a hierarchy of simulators is offered as a conceptual approach for assessing forest responses to changing climate and air quality across 13 southeastern states of the USA. This assessment approach combines geographic information system and Monte Carlo capabilities with several scales of computer modeling for southern pine species...

  19. Completeness and accuracy of data transfer of routine maternal health services data in the greater Accra region

    NARCIS (Netherlands)

    Amoakoh-Coleman, Mary; Kayode, Gbenga A.; Brown-Davies, Charles; Agyepong, Irene Akua; Grobbee, DE; Klipstein-Grobusch, Kerstin; Ansah, Evelyn K.

    2015-01-01

    Background: High quality routine health system data is essential for tracking progress towards attainment of the Millennium Development Goals 4 & 5. This study aimed to determine the completeness and accuracy of transfer of routine maternal health service data at health facility, district and

  20. 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)

  1. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    Energy Technology Data Exchange (ETDEWEB)

    J.S. Tang

    2001-05-03

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

  2. Gamma and Neutron Radiolysis in the 21-PWR Waste Package

    International Nuclear Information System (INIS)

    J.S. Tang

    2001-01-01

    The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

  3. Design and Development of Virtual Reactivity System for PWR

    International Nuclear Information System (INIS)

    Anwar, M. I.

    2012-01-01

    The reactivity monitoring and investigation is an important mean to ensure the safety operation of a nuclear power plant. But the reactivity of the nuclear reactor usually cannot be directly measured. It should be computed with certain estimation method. In this thesis, an effort has been made using an artificial neural network and highly fluctuating experimental data for predicting the total reactivity of the nuclear reactor based on all components of net reactivity. This virtual reactivity system is designed by taking advantage of neural network's nonlinear mapping capability. Based on analysis of the reactivity contributing factors, several neural network models are built separately for control rod, boron, poisons, fuel Doppler Effect and moderator effect. Extensive simulation and validation tests for PWR show that satisfied results have been obtained with the proposed approach. It presents a new idea to estimate the PWR's reactivity using artificial intelligence. All the design and simulation work is carried out in MATLAB and a real time programming environment is chosen for the computation and prediction of reactivity. (author)

  4. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  5. Liquid radioactive waste processing improvement of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes

    2005-01-01

    The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

  6. Initial Release of Nucliders from Spent PWR Fuels

    International Nuclear Information System (INIS)

    Kim, S. S.; Chun, K. S.; Kim, Y. B.; Choi, J. W.

    1994-01-01

    The relationship between the leaching and gap inventory of spent fuel has been studied. When a specimen of J44H08 spent PWR fuel with 38 GWD/MTU has been leached in the synthetic granitic groundwater in Ar atmosphere, the released fraction of cesium was increased rapidly up to 0,7% at around 500 days and stayed below 0.8% until 3 years. This 0.7% of cesium might be released from the gap in this fuel. The measurement of gap inventory with C15I08 spent PWR fuel, having 35 GWD/MTU and 0.22% of fission gas release, was also determined near 0.6% for the cesium, which is a similar fraction of cesium released from the leaching experiment with J44H08 fuel. Its gap inventories of strontium and iodine were about 0.03 and less than 0.2% respectively. Respective fractions of cesium and strontium in grain boundary of C15I08 were 0.78, 0.009%

  7. 21-PWR Waste Package Side and End Impacts

    International Nuclear Information System (INIS)

    T. Schmitt

    2005-01-01

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1

  8. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-10-26

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  9. Nucleotide sequence of the leading region adjacent to the origin of transfer on plasmid F and its conservation among conjugative plasmids.

    Science.gov (United States)

    Loh, S; Cram, D; Skurray, R

    1989-10-01

    The leading region of the Escherichia coli K12 F plasmid is the first segment of DNA to be transferred into the recipient cell during conjugal transfer. We report the nucleotide sequence of the 64.20-66.77F portion of the leading region immediately adjacent to the origin of transfer, oriT. The 2582 bp region encodes three open reading frames, ORF95, ORF169 and ORF273; the product of ORF273, is equivalent in size and map location to the 35 kDa protein, 6d, previously described (Cram et al. 1984). S1 nuclease analyses of mRNA transcripts have identified a potential promoter for ORF95 and ORF273 and indicated that these ORFs are transcribed as a single transcript; in contrast, ORF169 appears to be transcribed from two overlapping promoters on the complementary DNA strand. The products of ORF95 and ORF273 are mainly hydrophilic and are probably located in the cytoplasm. ORF273 shares some homology with DNA-binding proteins. There is a signal peptide sequence at the NH2-terminus of ORF169 and the mature form of ORF169 probably resides in the periplasm due to its hydrophilic nature. Both ORF273 and ORF169 are well conserved among conjugative F-like and a few non-F-like plasmids. On the other hand, ORF95 sequences are only present on some of these plasmids. Several primosome and integration host factor recognition sites are present implicating this region in DNA metabolism and/or replication functions.

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

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    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 requirements of

  11. Three-dimensional multislice spiral computed tomographic angiography: a potentially useful tool for safer free tissue transfer to complicated regions

    DEFF Research Database (Denmark)

    Demirtas, Yener; Cifci, Mehmet; Kelahmetoglu, Osman

    2009-01-01

    Three-dimensional multislice spiral computed tomographic angiography (3D-MSCTA) is a minimally invasive method of vascular mapping. The aim of this study was to evaluate the clinical usefulness of this imaging technique in delineating the recipient vessels for safer free tissue transfer to compli......Three-dimensional multislice spiral computed tomographic angiography (3D-MSCTA) is a minimally invasive method of vascular mapping. The aim of this study was to evaluate the clinical usefulness of this imaging technique in delineating the recipient vessels for safer free tissue transfer...... be kept in mind, especially inthe patients with peripheral vascular disease. 3D-MSCTA has the potential to replace digital subtraction angiography for planning of microvascular reconstructions and newer devices with higher resolutions will probably increase the reliability of this technique. (c) 2009...

  12. Validating a method for transferring social values of ecosystem services between public lands in the Rocky Mountain region

    Science.gov (United States)

    Sherrouse, Benson C.; Semmens, Darius J.

    2014-01-01

    With growing pressures on ecosystem services, social values attributed to them are increasingly important to land management decisions. Social values, defined here as perceived values the public ascribes to ecosystem services, particularly cultural services, are generally not accounted for through economic markets or considered alongside economic and ecological values in ecosystem service assessments. Social-values data can be elicited through public value and preference surveys; however, limitations prevent them from being regularly collected. These limitations led to our three study objectives: (1) demonstrate an approach for applying benefit transfer, a nonmarket-valuation method, to spatially explicit social values; (2) validate the approach; and (3) identify potential improvements. We applied Social Values for Ecosystem Services (SolVES) to survey data for three national forests in Colorado and Wyoming. Social-value maps and models were generated, describing relationships between the maps and various combinations of environmental variables. Models from each forest were used to estimate social-value maps for the other forests via benefit transfer. Model performance was evaluated relative to the locally derived models. Performance varied with the number and type of environmental variables used, as well as differences in the forests' physical and social contexts. Enhanced metadata and better social-context matching could improve model transferability.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-10-01

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

  14. Comparison of alpha decay and alpha transfer reactions in the lead region. [R matrix theory, absolute alpha widths, cross sections, 93 MeV

    Energy Technology Data Exchange (ETDEWEB)

    DeVries, R.M.

    1978-01-01

    Data were taken for five transitions in the lead region allowing a quantitative comparison with corresponding alpha-decay data via R-matrix theory using the same target + alpha nuclear potential. Good agreement between the absolute reduced widths determined from the two sets of data suggests that in transfer reactions, as in alpha decay, an alpha particle in its ground state is transferred in a one-step process. In a separate analysis, elastic and total reaction cross sections for the systems ..cap alpha.. + /sup 208/Pb, /sup 209/Bi were analyzed to obtain a limited set of potentials which, in turn, were used to calculate absolute alpha widths. Existing shell model calculations give ..gamma../sub ..cap alpha..//sup 2/ values three orders of magnitude smaller.

  15. PWR FLECHT SEASET 21-rod bundle flow blockage task. Task plan report. FLECHT SEASET Program report No. 5

    International Nuclear Information System (INIS)

    Hochreiter, L.E.; Basel, R.A.; Dennis, R.J.; Lee, N.; Massie, H.W. Jr.; Loftus, M.J.; Rosal, E.R.; Valkovic, M.M.

    1980-10-01

    This report presents a descriptive plan of tests for the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). This task will consist of forced and gravity reflooding tests utilizing electrical heater rods to simulate PWR nuclear core fuel rod arrays. All tests will be performed with a cosine axial power profile. These tests are planned to be used to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 161-rod flow blockage bundle tests

  16. CUPIDON. A code modelling the thermal and mechanical behaviour of a PWR fuel rod during a LOCA

    International Nuclear Information System (INIS)

    Chagrot, M.

    1978-12-01

    In the scope of the PHEBUS experimental programme to be performed in Cadarache on the behaviour of PWR fuel assemblies under loss of coolant accidental conditions, a computer code has been developed to help design the experimental rods and to contribute to the definition of the test runs. This code, dubbed CUPIDON, deals only with the thermal and mechanical behaviour of the rods as well as the oxidation of the cladding outside surface; it does not include any thermohydraulic subroutine. Rather, it is coupled with the RELAP code for providing necessary input data such as coolant temperatures and pressures and cladding-to-coolant heat transfer coefficients. It is restricted to a single, non-irradiated rod of short length as representing the PHEBUS experimental conditions. (author)

  17. Reference upper shelf fracture toughness properties of PWR pressure vessel materials: neutral/basic flux PWR submerged-arc welds

    International Nuclear Information System (INIS)

    Lidbury, D.P.G.

    1987-10-01

    A generic data base, relating to the upper shelf fracture toughness properties (O ≤ T ≤ 300 0 C) of pressurised water reactor (PWR) pressure vessel submerged-arc welds, deposited using neutral or basic fluxes, has been compiled and is presented in summary form within the main body of this report. A comparison with the A533B-1 plate and A508-3 forging data presented in the Second (1982) Report of the Light Water Reactor Study Group suggests the upper shelf fracture toughness properties of RPV submerged-arc welds metals are such that, over the temperature range appropriate to PWR plant operation: (i) initiation toughnesses are generally less than those associated with A533B-1/A508-3 base metals containing < 0.010 wt% S; (ii) enhanced toughnesses, corresponding to 2.0 mm stable crack extension, are comparable with those expected of A533B-1 plate materials containing < 0.010 wt% S. The information gathering exercise has also confirmed that upper shelf toughnesses associated with the use of basic or neutral fluxes are higher than those associated with the use of acidic fluxes. (author)

  18. U.S.-MEXICO TECHNOLOGY TRANSFER; BILATERAL TECHNICAL EXCHANGES FOR SUSTAINABLE ECONOMIC GROWTH IN THE BORDER REGION

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Richard, D., Dr.

    2007-10-01

    The U.S. Department of Energy (DOE) maintains a strong commitment to transfer the results of its science and technology programs to the private sector. The intent is to apply innovative and sometimes advanced technologies to address needs while simultaneously stimulating new commercial business opportunities. Such focused “technology transfer” was evident in the late 1990s as the results of DOE investments in environmental management technology development led to new tools for characterizing and remediating contaminated sites as well as handling and minimizing the generation of hazardous wastes. The Department’s Office of Environmental Management was attempting to reduce the cost, accelerate the schedule, and improve the efficacy of clean-up efforts in the nuclear weapons complex. It recognized that resulting technologies had broader world market applications and that their commercialization would further reduce costs and facilitate deployment of improved technology at DOE sites. DOE’s Albuquerque Operations Office (now part of the National Nuclear Security Administration) began in 1995 to build the foundation for a technology exchange program with Mexico. Initial sponsorship for this work was provided by the Department’s Office of Environmental Management. As part of this effort, Applied Sciences Laboratory, Inc. (ASL) was contracted by the DOE Albuquerque office to identify Mexico’s priority environmental management needs, identify and evaluate DOE-sponsored technologies as potential solutions for those needs, and coordinate these opportunities with decision makers from Mexico’s federal government. That work led to an improved understanding of many key environmental challenges that Mexico faces and the many opportunities to apply DOE’s technologies to help resolve them. The above results constituted, in large part, the foundation for an initial DOE-funded program to apply the Department’s technology base to help address some of Mexico

  19. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

  20. TRSM-a thermal-hydraulic real-time simulation model for PWR

    International Nuclear Information System (INIS)

    Zhou Weichang

    1997-01-01

    TRSM (a Thermal-hydraulic Real-time Simulation Model) has been developed for PWR real-time simulation and best-estimate prediction of normal operating and abnormal accident conditions. It is a non-equilibrium two phase flow thermal-hydraulic model based on five basic conservation equations. A drift flux model is used to account for the unequal velocities of liquid and gaseous mixture, with or without the presence of the noncondensibles. Critical flow models are applied for break flow and valve flow calculations. A 5-regime two phase heat convection model is applied for clad-to-coolant as well as fluid-to-tubing heat transfer. A rigorous reactor coolant pump model is used to calculate the pressure drop and rise for the suction and discharge ends with complete pump characteristics curves included. The TRSM model has been adapted in the full-scale training simulator of Qinshan Nuclear Power Plant 300 MW unit to simulate the thermal-hydraulic performance of the NSSS. The simulation results of a cold leg LOCA and a steam generator tube rupture (SGTR) accident are presented

  1. RCS natural circulation in a PWR station blackout accident--an application of NRC mechanistic codes

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    This paper discusses the phenomenon of reactor coolant system (RCS) natural circulation in a PWR station blackout accident with the loss of all AC power and auxiliary feedwater (the TMLB' accident). Existing and future studies performed for the industry and the Nuclear Regulatory Commission (NRC) are summarized in the paper. During the core uncovery and core melt period of the high-pressure TMLB' accident, multi-dimensional natural circulation of gas flow (steam and other gas such as hydrogen and fission products) is likely to exist in the uncovered core and the upper plenum above. Meanwhile, counter-current gas flow may also exist in the hot leg piping except during the opening of a power-operated relief valve (PORV) or safety relief valve (SRV) on the pressurizer. As a result, some of the core decay heat is transferred to the upper plenum structures and ex-vessel piping and components, and the RCS pressure boundary may be heated to high temperature to challenge structural integrity

  2. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  3. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Yoshida, Kazuo

    1979-05-01

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

  4. 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)

  5. Design and retrofit of radiation monitoring system for the PWR nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Tao; Xiong Guohua; Lang Yukai; Guo Wei

    2011-01-01

    Radiation monitoring system is important for the PWR nuclear power plant, and the research of design methods and principles for the radiation monitoring system can greatly improve the design ability of the system for PWR nuclear power plant, and reduce the risk of system retrofit. According to the Nuclear power plant regulations and design specifications, and taking the design and retrofit experience of the radiation monitoring system in Daya Bay Nuclear Power Plant into account, the general design principles and requirements of the radiation monitoring system in the PWR nuclear power plant is proposed, and the retrofit method of the radiation monitoring system in Daya Bay Nuclear Power Plant is briefly introduced. (authors)

  6. Status and future perspectives of PWR and comparing views on WWER fuel technology

    International Nuclear Information System (INIS)

    Weidinger, H.

    2003-01-01

    The main purpose of this paper is to give an overview on status and future perspectives of the Western PWR fuel technology. For easer understanding and correlating, some comparing views to the WWER fuel technology are provided. This overview of the PWR fuel technology of course can not go into the details of the today used designs of fuel, fuel rods and fuel assemblies. However, it tries to describe the today achieved capability of PWR fuel technology with regard to reliability, efficiency and safety

  7. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.

    1978-09-01

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  8. Isolation of probes specific to human chromosomal region 6p21 from immunoselected irradiation-fusion gene transfer hybrids

    International Nuclear Information System (INIS)

    Ragoussis, J.; Jones, T.A.; Sheer, D.; Shrimpton, A.E.; Goodfellow, P.N.; Trowsdale, J.; Ziegler, A.

    1991-01-01

    A hybrid cell line (R21/B1) containing a truncated human chromosome 6 (6pter-6q21) and a human Y chromosome on a hamster background was irradiated and fused to A23 (TK-) or W3GH (HPRT-) hamster cells. Clones containing expressed HLA class I genes (4/40) were selected using monoclonal antibodies. These clones were recloned and analyzed with a panel of probes from the HLA region. One hybrid (4G6) contained the entire HLA complex. Two other hybrids (4J4 and 4H2) contained only the HLA class I region, while the fourth hybrid (5P9) contained HLA class I and III genes in addition to other genes located in the 6p21 chromosomal region. In situ hybridization showed that the hybrid cells contained more than one fragment of human DNA. Alu and LINE PCR products were derived from these cells and compared to each other as well as to products from two somatic cell hybrids having the 6p21 region in common. The PCR fragments were then screened on conventional Southern blots of the somatic cell hybrids to select a panel of novel probes encompassing the 6p21 region. In addition, the origin of the human DNA fragments in hybrid 4J4 was determined by regional mapping of PCR products

  9. Isolation of probes specific to human chromosomal region 6p21 from immunoselected irradiation-fusion gene transfer hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Ragoussis, J.; Jones, T.A.; Sheer, D.; Shrimpton, A.E.; Goodfellow, P.N.; Trowsdale, J.; Ziegler, A. (ICRF Human Immunogenetics, London (England))

    1991-07-01

    A hybrid cell line (R21/B1) containing a truncated human chromosome 6 (6pter-6q21) and a human Y chromosome on a hamster background was irradiated and fused to A23 (TK-) or W3GH (HPRT-) hamster cells. Clones containing expressed HLA class I genes (4/40) were selected using monoclonal antibodies. These clones were recloned and analyzed with a panel of probes from the HLA region. One hybrid (4G6) contained the entire HLA complex. Two other hybrids (4J4 and 4H2) contained only the HLA class I region, while the fourth hybrid (5P9) contained HLA class I and III genes in addition to other genes located in the 6p21 chromosomal region. In situ hybridization showed that the hybrid cells contained more than one fragment of human DNA. Alu and LINE PCR products were derived from these cells and compared to each other as well as to products from two somatic cell hybrids having the 6p21 region in common. The PCR fragments were then screened on conventional Southern blots of the somatic cell hybrids to select a panel of novel probes encompassing the 6p21 region. In addition, the origin of the human DNA fragments in hybrid 4J4 was determined by regional mapping of PCR products.

  10. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  11. Atmospheric Transference of the Toxic Burden of Atmosphere-Surface Exchangeable Pollutants to the Great Lakes Region

    Science.gov (United States)

    Kumar, A.; Perlinger, J. A.; Giang, A.; Zhang, H.; Selin, N. E.; Wu, S.

    2016-12-01

    Toxic pollutants that share certain chemical properties undergo repeated emission and deposition between Earth's surfaces and the atmosphere. Following their emission through anthropogenic activities, they are transported locally, regionally or globally through the atmosphere, are deposited, and impact local ecosystems, in some cases as a result of bioaccumulation in food webs. We call them atmosphere-surface exchangeable pollutants or "ASEPs", wherein this group is comprised of thousands of chemicals. We are studying potential future contamination in the Great Lakes region by modeling scenarios of the future for three compounds/compound classes, mercury, polychlorinated biphenyl compounds, and polycyclic aromatic hydrocarbons. In this presentation we focus on mercury and future scenarios of contamination of the Great Lake region. The atmospheric transport of mercury under specific scenarios will be discussed. The global 3-D chemical transport model GEOS-Chem has been applied to estimate future atmospheric concentrations and deposition rates of mercury in the Great Lakes region for selected future scenarios of emissions and climate. We find that, assuming no changes in climate, annual mean net deposition flux of mercury to the Great Lakes Region may increase by approximately 50% over 2005 levels by 2050, without global or regional policies addressing mercury, air pollution, and climate. In contrast, we project that the combination of global and North American action on mercury could lead to a 21% reduction in deposition from 2005 levels by 2050. US action alone results in a projected 18% reduction over 2005 levels by 2050. We also find that, assuming no changes in anthropogenic emissions, climate change and biomass burning emissions would, respectively, cause annual mean net deposition flux of mercury to the Great Lakes Region to increase by approximately 5% and decrease by approximately 2% over 2000 levels by 2050.

  12. Incorporation of an evolutionary algorithm to estimate transfer-functions for a parameter regionalization scheme of a rainfall-runoff model

    Science.gov (United States)

    Klotz, Daniel; Herrnegger, Mathew; Schulz, Karsten

    2016-04-01

    This contribution presents a framework, which enables the use of an Evolutionary Algorithm (EA) for the calibration and regionalization of the hydrological model COSEROreg. COSEROreg uses an updated version of the HBV-type model COSERO (Kling et al. 2014) for the modelling of hydrological processes and is embedded in a parameter regionalization scheme based on Samaniego et al. (2010). The latter uses subscale-information to estimate model via a-priori chosen transfer functions (often derived from pedotransfer functions). However, the transferability of the regionalization scheme to different model-concepts and the integration of new forms of subscale information is not straightforward. (i) The usefulness of (new) single sub-scale information layers is unknown beforehand. (ii) Additionally, the establishment of functional relationships between these (possibly meaningless) sub-scale information layers and the distributed model parameters remain a central challenge in the implementation of a regionalization procedure. The proposed method theoretically provides a framework to overcome this challenge. The implementation of the EA encompasses the following procedure: First, a formal grammar is specified (Ryan et al., 1998). The construction of the grammar thereby defines the set of possible transfer functions and also allows to incorporate hydrological domain knowledge into the search itself. The EA iterates over the given space by combining parameterized basic functions (e.g. linear- or exponential functions) and sub-scale information layers into transfer functions, which are then used in COSEROreg. However, a pre-selection model is applied beforehand to sort out unfeasible proposals by the EA and to reduce the necessary model runs. A second optimization routine is used to optimize the parameters of the transfer functions proposed by the EA. This concept, namely using two nested optimization loops, is inspired by the idea of Lamarckian Evolution and Baldwin Effect

  13. Sizewell 'B' PWR pre-construction safety report

    International Nuclear Information System (INIS)

    1982-04-01

    The Pre-Construction Safety Report (PCSR) for a PWR power station to be constructed as Sizewell 'B' is presented in 13 volumes containing 16 chapters. The PCSR has been submitted to the Nuclear Installations Inspectorate in support of the Central Electricity Generating Board's application for consent to the extension at Sizewell. It describes the design and provides the safety case for the proposed station, which comprises a 4-loop pressurized water reactor with associated generating plant and supporting auxiliary equipment. A general description of the station and its site is given. The strategy for ensuring nuclear safety is set out and the general design aspects of systems and plant outlined. The plant and systems, including their safety design bases and the fault analyses carried out for the design are described. Finally the way in which the plant will be decommissioned at the end of its useful life is outlined. (U.K.)

  14. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-02-01

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented [fr

  15. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  16. Neutron measurements in borated water for PWR fuel inspections

    International Nuclear Information System (INIS)

    Rinard, P.M.

    1984-07-01

    A fork detector has been developed for use in the international effort to safeguard irradiated fuel assemblies. To improve interpretation of data from a fork, the following three facets of the detector's neutron counting response have been examined using a tank of borated water and a PWR fresh-fuel assembly: (1) The detector's sensitivity to neutrons initiated at different positions within the assembly was measured and this sensitivity can be used to generate total responses to assemblies with uniform or nonuniform irradiation. (2) Using fission chambers with and without cadmium wrappings provided ratios of count rates that can give an independent estimate of the boron concentration in the water. The precision of a boron determination can be estimated from these measurements. (3) The water temperature was raised, causing small but possibly important effects on the count rates. These facets of the fork detector's neutron response were studied at boron concentrations ranging from 0 to about 3500 ppM

  17. B ampersand W PWR advanced control system algorithm development

    International Nuclear Information System (INIS)

    Winks, R.W.; Wilson, T.L.; Amick, M.

    1992-01-01

    This paper discusses algorithm development of an Advanced Control System for the B ampersand W Pressurized Water Reactor (PWR) nuclear power plant. The paper summarizes the history of the project, describes the operation of the algorithm, and presents transient results from a simulation of the plant and control system. The history discusses the steps in the development process and the roles played by the utility owners, B ampersand W Nuclear Service Company (BWNS), Oak Ridge National Laboratory (ORNL), and the Foxboro Company. The algorithm description is a brief overview of the features of the control system. The transient results show that operation of the algorithm in a normal power maneuvering mode and in a moderately large upset following a feedwater pump trip

  18. Specification of water quality for the FRAMATOME PWR secondary circuit

    International Nuclear Information System (INIS)

    Nordmann, F.

    1980-03-01

    This paper describes the purpose, theory and scope of secondary system chemical specifications for FRAMATOME PWR nuclear power plants. All volatile treatment was chosen: controlling the feedwater pH by means of a volatile amine (ammonia, morpholine), and excluding oxygen by the addition of hydrazine. The pollutants are monitored at the steam generator drains by completely automatic measurements using simple and reliable techniques: pH measurement and a diagram of the cation conductivity versus sodium. An explanation is given of the monitoring techniques and to the effect of the various kinds of possible pollutant. A new concept is described, the annual quota expressed in day.microsiements.cm -1 which enables the amount of absorbed pollutants in the steam generator to be evaluated. The methods used for maintaining the desired chemical quality are dealt with [fr

  19. Reliability assessment of the containment of a PWR

    International Nuclear Information System (INIS)

    Schueller, G.I.; Wellein, R.; Wittmann, F.H.; Boulahdour, T.; Mihashi, M.; Zorn, N.F.; Bauer, J.

    1981-09-01

    The aim of this research effort was to contribute to the development of methods to quantify the risk involved with nuclear power plants. Using a large component, i.e. the containment of the reference plant BIBLIS B (PWR) as sample structure a reliability analysis was performed which is based on realistic assumptions of loads and material properties. For this purpose in many fields it was necessary to develop new methods, collect data, and where not available, obtain data in tests. This effort concentrated on partial aspects and on the other hand on the development of a methodology for an overall reliability concept. According to the results of the previous project, the keypoints of this effort are the treatment of loss of coolant accidents (small leak), earthquake loading, the possibly resulting crackpropagation in the steel hull, and the structural mechanics and material strength aspects of the reinforced concrete hull subjected to impact loading (aircraft impact). (orig./HP) [de

  20. Neutronal aspects of PWR control for transient load following

    International Nuclear Information System (INIS)

    Cossic, A.

    1985-01-01

    The purpose of this thesis is to qualify the CRONOS diffusion code on a load transient in grey mode control. First of all, we have established a general axial calculational model and studied the important physical phenomena: xenon oscillation, grey rods absorption, radial leaks modelling, effect of the initial conditions in Iodine and Xenon. In a second stage, a three dimensional calculation has been performed, the results of which have been compared to a PWR 900 TRICASTIN 3 experiment and have been in good agreement. In the last part, we show that the results of the axial model using one-dimensional CRONOS calculations are quite consistent with the three-dimensional calculation [fr

  1. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    Costaz, J.L.

    1987-01-01

    The analysis of the large amount of measurements which has been now gathered by EDF on its twenty two PWR 900 MW shows that the behaviour of concrete under creep and shrinkage effects is in good agreement with the values given as correct estimates by french regulations and taken into account for the design of nuclear prestressed structures. None of the containment buildings studied here showed significant differences with the regulations theoretical values and consequently all the measurements remain in the field of the allowable strain variations used for design. On the other hand, if the instant loading elastic modulus is clearly determined for each containment, and its effect on theoretical creep taken into account, it was not possible up till now to extract from measurements some particular effects such as type of concrete and agregates or climatic effects. (orig.)

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

    International Nuclear Information System (INIS)

    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

  3. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    Poncet, B.

    1982-11-01

    For the precise industrial purpose of PWR nuclear power plant steam generator water level control, a natural method is developed where classical techniques seem not to be efficient enough. From this essentially non-linear practical problem, an input-output identification of dynamic systems is proposed. Through Homodynamic Systems, characterized by a regularity property which can be found in most industrial processes with balance set, state form realizations are built, which resolve the exact joining of local dynamic behaviors, in both discrete and continuous time cases, avoiding any load parameter. Specifically non-linear modelling analytical means, which have no influence on local joined behaviors, are also pointed out. Non-linear autoregressive realizations allow us to perform indirect adaptive control under constraint of an admissible given dynamic family [fr

  4. Optimization of the decontamination in EDF PWR power plants

    International Nuclear Information System (INIS)

    Gosset, P.; Dupin, M.; Buisine, D.; Buet, J.F.; Brunel, V.

    2002-01-01

    The optimisation of decontamination in EDF PWR power plants is the result of a permanent collaborative work between the plant operators, the subcontractors, central services of nuclear power division of EDF. This collaborative work enables the saving of all the feedback experience. The main operations carried out on nuclear sites like mechanical decontamination of valves, use of the ''EMMAC'' process on big components (replacement of steam generator, hydraulic parts of the reactor coolant pumps), use of foam on pools walls and divers in highly contaminated pools have been discussed. This paper shows that the choice of decontamination processes is very dependant on the components, on the dose rate reduction to be aimed and on the possibility to treat the waste on site. (authors)

  5. 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.

  6. Low concentration NP preoxidation condition for PWR decontamination

    International Nuclear Information System (INIS)

    Huang Fuduan; Yu Degui; Lu Jingju; Ding Dejun; Zhao Yukun

    1991-02-01

    To use preoxidation condition with low concentration NP (nitric acid permanganate) instead of conventional high concentration AP (alkline permanganate ) for PWR oxidation decontamination (POD) was summarized. Experiments including three parts have been performed. The defilming performance and decontamination factor of preoxidation with low concentration NP, which is 100, 10 times lower than that of AP are better than that with high concentration AP. The reason has been studied with the aid of prefilmed specimens of corrosion potential measuring in NP solution and chromium release in NP and AP solutions. The behaviour of alloy 13 prefilmed specimen in NP preoxidation solution is different from 18-8 ss and Incoloy 800. In the low acidity, the corrosion potential moves toward positive direction as the acidity becomes high

  7. Contribution to the experimental qualification of PWR fuel storage calculations

    International Nuclear Information System (INIS)

    Marsault, Philippe.

    1980-12-01

    Experiments were carried out on assemblies representative of those used in PWR reactors in a configuration made critical with a driver zone. In this way, certain parameters were able to be measured using current classical techniques. As the multiplication factor for a group of assemblies cannot be determined directly, substitutions were made with an equivalent homogeneous lattice in which Laplacian measurements could be made. The k(infinite) factor was obtained by introducing a migration area which can only be obtained from calculations. Experimental storage studies realized during the CRISTO 1 campaign utilize: 1) a lattice with 4 14x14 pin assemblies immersed in ordinary water; 2) a lattice with 4 14x14 pin assemblies and 3) a regular lattice. The CRISTO experiment enabled criticality calculations to be qualified with these lattices for storage under accidental conditions [fr

  8. 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.

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

    International Nuclear Information System (INIS)

    Rouf; Su'ud, Zaki

    2016-01-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O 2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd 2 O 3 ) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively. (paper)

  10. The perils of technology transfer : the Australian wheat/medic System in the Near East/North Africa region

    Directory of Open Access Journals (Sweden)

    Risopoulos, S.

    1990-01-01

    Full Text Available Yields and production of rainfed areas in the Near East and North Africa are stagnating. The Australian wheat-medic system has been tried out in several countries of the region, Increases in soil fertility and yields were expected as well as better crop-livestock integration. Difficulties were more serious than foreseen. The farmer of the region differs from his Australian counterpart by the much smaller size of his farm and by his preference for keeping his land-use options open to match climatic variability.

  11. Radiation risk analysis of tritium in PWR plants

    International Nuclear Information System (INIS)

    Yang Maochun; Wang Shimin

    1999-03-01

    Tritium is a common radionuclide in PWR nuclear power plant. In the normal operation conditions, its radiation risk to plant workers is the internal radiation exposure when tritium existing in air as HTO (hydrogen tritium oxide) is breathed in. As the HTO has the same physical and chemical characteristics as water, the main way that HTO entering the air is by evaporation. There are few opening systems in Nuclear Power Plant, the radiation risk of tritium mainly exists near the area of spent fuel pit and reactor pit. The highest possible radiation risk it may cause--the maximum concentration in air is the level when equilibrium is established between water and air phases for tritium. The author analyzed the relationship among the concentration of HTO in water, in air and the water temperature when equilibrium is established, the equilibrated HTO concentration in air increases with HTO concentration in water and water temperature. The analysis revealed that at 30 degree C, the equilibrated HTO concentration in air might reach 1 DAC (derived air concentration) when the HTO concentration in water is 28 GBq/m 3 . Owing to the operation of plant ventilation systems and the existence of moisture in the input air of the ventilation, the practical tritium concentration in air is much lower than its equilibrated levels, the radiation risk of tritium in PWR plant is quite limited. In 1997, Daya Bay Nuclear Power Plant's practical monitoring result of the HTO concentration in the air of the nuclear island and the urine of workers supported this conclusion. Based on this analysis, some suggestions to the reduction of tritium radiation risk were made

  12. Probe for detection of denting in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Germain, J.L.; Nio, J.C.

    1994-07-01

    In certain types of PWR steam generator, oxide deposits can lead to embedding, and subsequently to deformation of a tube (the phenomenon of ''denting''). Such embedding changes the vibratory behavior of the tubes and can result in fatigue cracking. This type of cracking can also be worsened in the event of improper assembly of the anti-vibration spacer bars supporting the U-bends. To prevent such incidents and provide for effective preventive condition-directed maintenance of its PWR steam generators, EDF has undertaken the study and development of a probe to detect this type of phenomenon. The studies began in 1990 and led to the building of an initial prototype probe. The principle behind the probe consists in inducing vibration in the U-bend and determining the main resonance modes of the tube. Measurements of frequency and amplitude and calculation of damping enable characterization of the mechanical behavior of the U-bend. The most important parameter is damping, for which the value must be sufficiently high to ensure that the tube is not subjected to major vibratory amplitudes during operation. Numerous tests have been performed with the first prototype version of the probe, on a mock-up in the test area and on one of the demounted steam generators on the Dampierre site. These different tests have enabled validation of the operating principle, fine-tuning the process, pinpointing certain mechanical problems in the probe design, and obtaining the first indications as to the real vibratory behavior of U-bends on a steam generator. On the basis of these preliminary tests, the specifications were drawn up for an industrial version of the probe. Following a call for bids and the choice of a manufacturer, work began on fabrication of a new probe model in 1993. This version was delivered at the end of 1993 and testing began in 1994. (authors). 5 figs., 2 tabs

  13. Corrosion Resistance Evaluation of HANA Claddings in Commercial PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Hun; Kwon, Oh-Hyun; Kim, Hong-Jin; Yoo, Jong-Sung; Kim, Yong-Hwan [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    Korea Atomic Energy Research Institute (KAERI) in collaboration with KEPCO Nuclear Fuel (KNF) developed newly-advanced alloy which are named HANA (High-performance Alloy for Nuclear Application) for high burnup PWR nuclear fuel, showed an excellent out-pile corrosion resistance in PWR simulating loop conditions. And in-pile corrosion resistance of HANA claddings, which was examined at the first provisional inspection after -185 FPD of irradiation in the Halden Reactor, and also shown superior to the other references alloy. Also, other researches showed a much better corrosion resistance when compared to the other Zr-based alloy in various corrosion conditions. In this study, the LTA program for newly-developed fuel assembly (HIPER) with the HANA claddings was implemented to justify the performance for 3 cycles of operation schedule in Hanul nuclear power plant. The objective of this study is to compare corrosion properties of reference alloy with HANA claddings loaded in Hanul nuclear power plant.. For the examination procedures, the oxide thickness measurements method and equipment of PSE are described in detail as follow in measurement methods chapter. Finally, based on the above mentioned measurements method, the summarized oxide thickness data obtained from PSE are evaluated for the corrosion resistance in commercial nuclear power plant and some discussion for the corrosion resistance are described. In the past, corrosion resistance of HANA claddings was successfully conducted in test reactor. In this study, the corrosion characteristic of HANA claddings which are applied to HIPER is examined in the commercial nuclear power plant. HANA claddings in the HIPER showed a more improved corrosion resistance than reference alloy claddings and are evaluated well with meeting the oxide thickness criteria.

  14. 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.

  15. Inelastic electron scattering from 3He and 4He in the threshold region at high momentum transfer

    International Nuclear Information System (INIS)

    Rock, S.; Arnold, R.G.; Chertok, B.T.; Szalata, Z.M.; Day, D.; McCarthy, J.S.; Martin, F.; Mecking, B.A.; Sick, I.; Tamas, G.

    1981-01-01

    The cross section for inclusive inelastic electron scattering from the helium isotopes has been measured at momentum transfers squared of 0.8 less than or equal to Q 2 less than or equal to 5.0 (GeV/c) 2 for 3 He and 0.8 less than or equal to Q 2 less than or equal to 2.4 (GeV/c) 2 for 4 He. The data were taken at 10 0 and cover the range 1.0 2 /2M/sub He/ν, which includes the elastic peak, nuclear breakup threshold, the high momentum tail of the quasi elastic scattering, and pion production. The structure function, νW 2 , derived from the data is approaching a scaling limit at high Q 2 . It can be factored into a product of functions of Q 2 and of x as predicted by some models

  16. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    International Nuclear Information System (INIS)

    Kavaklioglu, K.; Ikonomopoulos, A.

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

  17. Coordinated U. S. PWR Reactor Vessel Surveillance Program: Surveillance Data to Support Long Term Operations

    International Nuclear Information System (INIS)

    Hosler, Ryan; Troyer, Greg; Davidsaver, Sarah; Hardin, Timothy

    2012-01-01

    Irradiated reactor pressure vessel (RPV) surveillance data is used as the basis for embrittlement trend correlations (ETCs) which predict decreases in RP fracture toughness due to irradiation embrittlement. A limited amount of data exists today at fluences that many U. S. PWR RPVs will reach in 60 or more years of operation. However, there is a significant amount of test reactor data available at high fluences, which shows higher embrittlement shifts than the power reactor data-based correlations. A coordinated plan for withdrawal and testing of the U. S. PWR RPV surveillance capsules has been developed, with the intent of filling high fluence gaps in existing PWR data. This paper summarizes the methodology, optimization strategy, and current results of this coordinated U. S. PWR reactor vessel surveillance program (CRVSP). The Coordinated RVSP has been optimized to maximize the quantity and quality of high fluence data while minimizing the burden on the industry

  18. A method to determine the dampening system of control rod drop mechanism for PWR reactors

    International Nuclear Information System (INIS)

    Trindade, C.E.; Mattos, J.R.L. de; Perrotta, J.A.

    1988-08-01

    A method to determine the Control Assembly damping drop system (dashpot/guide tube) was developed. It's presented a theoretical model, an experimental device and the procedures to determine this system, which is used in PWR reactors. (author) [pt

  19. Aspects of PWR nuclear power plant secondary cycle relating to reactor safety

    International Nuclear Information System (INIS)

    Mueller, A.E.F.; Leal, M.R.L.V.; Dominguez, D.

    1981-01-01

    A safety study of the main steam system, condensate and feedwater systems and water treatment system that belong to the secondary cooling circuits of a PWR nuclear power plant is presented. (E.G.) [pt

  20. EDF's PWR power plants: anomalies concerning the reactor core instrumentation system

    International Nuclear Information System (INIS)

    1985-10-01

    This report presents the problems of fatigue and leaks found on the internal core instrumentation thimbles of several French PWR power plants, as also the solutions chosen according the reactor has already or not been operating [fr

  1. Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

    International Nuclear Information System (INIS)

    Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

    2001-01-01

    Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

  2. Computational study of heat transfer from the inner surface of a circular tube to force high temperature liquid metal flow in laminar and transition regions

    Science.gov (United States)

    Hata, K.; Fukuda, K.; Masuzaki, S.

    2018-03-01

    Heat transfer through forced convection from the inner surface of a circular tube to force the flow of liquid sodium in the laminar and transition regions were numerically analysed for two types of tube geometries (concentric annular and circular tubes) and two types of equivalent diameters (hydraulic and thermal equivalent diameters). The unsteady laminar three-dimensional basic equations for forced convection heat transfer caused by a step heat flux were numerically solved until a steady state is attained. The code of the parabolic hyperbolic or elliptic numerical integration code series (PHOENICS) was used for calculations by considering relevant temperature dependent thermo-physical properties. The concentric annular tube has a test tube with inner and outer diameters of 7.6 and 14.3 mm, respectively, has a heated length of 52 mm, and an L/d of 6.84. The two circular tubes have inner diameters of 6.7 and 19.3 mm with L/d of 7.76 and 2.69, respectively, and a heated length of 52 mm. The inlet liquid temperature, inlet liquid velocity, and surface heat flux were equally set for each test tube as T in ≅573 to 585 K, u in = 0.0852 to 1 m/s, and q = 2×105 to 2.5×106 W/m2, respectively. The increase in temperature from the leading edge of the heated section to the outlet of the circular tubes (with a hydraulic diameter of d H = 6.7 mm and a thermal equivalent diameter d te = 19.3 mm) was approximately 2.70 and 1.21 times as large as the corresponding values of the concentric annular tube with an inner diameter of 7.6 mm and an outer diameter of 14.3 mm, respectively. A quantity in the laminar and transition regions was suggested as the dominant variable involved in the forced convection heat transfer in the circular tube. The values of the local and average Nusselt numbers, Nu z and Nu av , respectively, for a concentric annular tube with d H = 6.7 mm and for a circular tube with d H = 6.7 mm were calculated to examine the effects of q, T in , and Pe on heat

  3. CO2 concentration characteristics and possible influence of waves on the rate of CO2 transfer between the ocean and the atmosphere in a coastal region.

    Science.gov (United States)

    Herrera-Vazquez, Carlos F.; Ocampo-Torres, Francisco J.

    2017-04-01

    In order to understand the physical processes involved in the air-sea transfer velocity of CO2 in a coastal region. The possible influence of the waves as an external agent is studied in order to characterize the CO2 transfer. The air-sea transfer velocity of CO2 was calculated from direct measurements of CO2 flux and CO2 partial pressure difference at the area of Punta Morro in Ensenada, B. C., Mexico during the period from 13 April to 3 May of 2016. CO2 fluxes were measured at the coastline at a height of 10m by a flux measurement tower using eddy covariance method; in the sea, at a distance of approximately 1000m from the measuring tower, a CO2 sensor (Pro-Oceanus) was used to measure the CO2 partial pressures in air and sea water at a distance of approximately 2m of the surface. On the sea bottom at a depth of 10m and 400m from the coastline, a CO2 sensor (SAMI-CO2) and acoustic profiler (Aquadopp, Nortek AS) were installed measuring CO2 partial pressure in the sea water and waves, respectively. The results show that CO2 concentration is not homogeneous in the study area, we were able to identify both horizontal and vertical gradients of pCO2 in the air and in sea water. Close to the sea surface, values of pCO2 in sea water were always smaller than there in air. The measured CO2 flux was in average negative during our field experiment. The air-sea transfer velocity of CO2 was obtained, resulting in a subtle relation with the significant wave height incident to the coast.This work is a RugDiSMar project (CONACYT 155793) contribution. Partial support from CB-2015-01-255377 is appreciated.

  4. Changing characteristics of land use and ecological service value in the water source region of the Middle Route of South-to-North Water Transfer Project

    Science.gov (United States)

    Tang, Jian; Zhai, Wenliang; Cao, Huiqun

    2017-08-01

    Research on changing characteristics of land use and ecological service value (ESV) can guide the regional land use planning and promote the rational use of environmental resources. On the basis of four phases of land-use data (2000, 2005, 2010 and 2015), this research analysed the changing characteristics of land use and ESV in the water source region of the Middle Route of South-to-North Water Transfer Project (SRMRP). The results showed that forest, grassland and cultivated land were the major land-use types in the SRMRP. During 2000∼2015, forest, grassland, farmland and wetland decreased. Construction land and bare land had increased, and the annual increase rates reached 3.6% and 8%, respectively. After the implementation of the water transfer project in 2003, water area was also increasing. The total ESV in the SRMRP is about 196 billion CNY, and mainly comes from the contributions of forest, grassland and farmland. During 2000∼2015, farmland shrinks leaded to the declines in value from supply service. With increasing in water and construction land, value from entertainment and cultural service increased. During the early stage of the water transfer project, value from regulation and support services increased due to the increase in water. With the decreasing in wetland and the increasing in construction land, the negative effects on the regulation and support services were increasing, and value from regulation and support services were therefore decreasing. During the process of resource exploitation and management, more attentions should be paid to the total control of construction land and wetland protection in the SRMRP.

  5. Resfria - a computational routine for thermal-hydraulic analysis of a cooldown in the PWR

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Maciel Filho, L.A.

    1989-01-01

    This paper presents the computer code RESFRIA, designed to calculate the process parameters in a PWR nuclear power plant during a cooldown normal procedure. The procedure is described and some of the models developed to the simulation of systems and equipments are presented. A simplified flowchart of the computational routine and the results in the form of a diagram, for a typical PWR nuclear power plant, are also presented. (author)

  6. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.; Fero, A.; Snyder, M.

    2004-01-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR

  7. Teknologi Pembuatan Cermet Du0¬2 - Steel Untuk Wadah Limbah Bahan Bakar Bekas Pwr

    OpenAIRE

    Alimah, Siti; Budiarto, Budiarto

    2005-01-01

    DUO­2-STEEL CERMET MANUFACTURING TECHNOLOGY FOR PWR Spent Nuclear Fuel (SNF) CASKS. Assessment of DU02 - Steel cermet manufacturing technology for PWR SNF casks has been done. DU02 - Steel cermet consisting of DU02 particulates and other particulates, embedded in a steel matrix. Cermet SNF casks have the potential for superior performance compared with casks constructed of other materials. The addition of DU02 ceramic particulates can increase SNF cask capacity, improve of repository performa...

  8. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    International Nuclear Information System (INIS)

    Tylee, J.L.

    1981-01-01

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method

  9. Calibration of four neutron coincidence collars for PWR fresh fuel assemblies

    International Nuclear Information System (INIS)

    De Baere, P.; Carchon, R.; Smaers, G.; Smith, B.G.R.; Cranston, R.; Levy-Gorget, J.L.

    1988-05-01

    A measurement campaign was set up in order to calibrate four Neutron Coincidence Collars. For this purpose, a PWR fuel mock-up was used, as well as a series of real size PWR fuel assemblies. Calibration functions were set up, representing net real coincidence rate as a function of mass loading. All these calibration expressions have been referred to a general calibration expression, by applying some correction factors on the real coincidence count rate. (Author)

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

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.

    2004-01-01

    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

  11. 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.

  12. Dry Ice Blast Decontamination to in-service equipment in Japanese PWR plant

    International Nuclear Information System (INIS)

    2016-01-01

    MHI had developed several mechanical decontamination methods. Mechanical decontamination is beneficial when it is applied to equipment whose surface is narrow. Especially in terms of secondary waste reduction, MHI started the study of application of Dry Ice Blast Decontamination to actual PWR plant. This paper provides an introduction to Dry Ice Blast Decontamination principle, its system and actual application result to PWR plant. (J.P.N.)

  13. Tn7 and Tn501 Insertions into Pseudomonas aeruginosa plasmid R91-5: mapping of two transfer regions.

    OpenAIRE

    Moore, R J; Krishnapillai, V

    1982-01-01

    We constructed a restriction endonuclease map of the Pseudomonas aeruginosa narrow-host-range plasmid R91-5. Insertions of transposons Tn7 and Tn501 into the plasmid DNA were characterized physically and genetically. The distribution of sites of insertion showed some regional specificity for the insertion of these transposons, especially TN501. The insertion of Tn7 was unusual in that all 42 of 43 insertions were in the same orientation. By relating phenotypic changes to the site of insertion...

  14. Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

    International Nuclear Information System (INIS)

    Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

    2017-01-01

    In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

  15. Time trends (1986-2003) of radiocesium transfer to roe deer and wild boar in two Austrian forest regions

    Energy Technology Data Exchange (ETDEWEB)

    Strebl, F. [Austrian Research Centers GmbH - ARC, Radiation Safety and Applications, A-2444 Seibersdorf (Austria)], E-mail: friederike.strebl@arcs.ac.at; Tataruch, F. [Research Institute of Wildlife Ecology, University of Veterinary Medicine, Savoyenstr.1, A 1160 Vienna (Austria)

    2007-11-15

    Starting shortly after the Chernobyl accident, samples of roe deer and wild boar from two comparatively highly contaminated Austrian forest stands have been regularly analysed for {sup 137}Cs. Until 1995 average {sup 137}Cs concentrations exceeded 1000 Bq kg{sup -1} in both roe deer and wild boar. Long-term and seasonal trends are similar in both investigation sites. While {sup 137}Cs aggregated transfer factor (T{sub ag}) values show a significant decreasing trend in roe deer (ecological half-time 8.6 and 7.2 years, respectively), T{sub ag}-values in wild boar are highly variable, but rather increasing values are observed over the last years. T{sub ag}-values for roe deer are between 0.04 and 0.008 m{sup 2} kg{sup -1} fresh weight (1987-2003); values for wild boar are between 0.008 m{sup 2} kg{sup -1} (1988) and 0.046 m{sup 2} kg{sup -1} (1996) fresh weight. Seasonal trends for both species are in good agreement with observations from German forests: increased mushroom ingestion leads to higher {sup 137}Cs T{sub ag}-values for roe deer in the second half of the year (August-December) compared to the first half (January-July). T{sub ag}-values for wild boar are highest in the first half of the year.

  16. The contribution of multidimensional spatial analysis to a waste management policy: implementation of the ELECTRE method for characterizing transfer centers in the region of Oran

    Science.gov (United States)

    Saidi, A.; Trache, M. A.; Khelfi, M. F.

    2016-08-01

    The social and economic activity steadily growing in our cities creates a significant waste production in constantly evolving. The management of this waste is problematic because it is the center of many issues and interests. Indeed, any action or decision to the collection, transportation, treatment and disposal of waste should be considered in the economic, social, political and especially environmental aspect. A global Geomatic solution requires implementing a GIS with powerful multidimensional spatial analysis tools that support really waste management problem. Algeria has adopted a solution of waste landfill for all urban cities. In the Oran region, it exists three Centers Controlled landfill (CET) which the most important is that of Hassi-Bounif. This center currently meeting the needs of the region is unsustainable solution at the long-term because of its rapid saturation and its geographic location, which is still far from city centers (20-30 km) implying a negative impact on the vehicle park collecting such frequent breakdowns, the rapid degradation, slow delivery time and especially the high cost of the maintenance operation. This phenomenon is aggravated by the absence of real and actual initiatives targeting the recycling and recovery of waste, which makes the CET an endpoint for all types of waste. We present in this study, the use of the ELECTRE method (Multicriteria Analysis) integrated into a GIS to characterize the impact of the implementation of transfers centers at Oran region. The results of this study will accentuate the advantages of the activation of waste warehouse closer to the city, and relieving considerably the volume of transfer towards CET. The objective of our presentation is to show the leading role of the new Geomatics tools and the multidimensional spatial analysis in the apprehension of an environmental problem such the waste management and more generally in the urban management.

  17. Thermal Design Feasibility of Th-{sup 233}U PWR Breeder

    Energy Technology Data Exchange (ETDEWEB)

    Volasky, Daphna; Shwageraus, Eugene [Department of Nuclear Engineering, Ben-Gurion University of the Negev, Beer Sheva 84105 (Israel); Fridman, Emil [Forschungszentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Dresden 01314 (Germany)

    2009-06-15

    The Light Water Breeder Reactor program (LWBR) proved that a sustainable closed fuel cycle based on reprocessed uranium {sup 233}U and {sup 232}Th feed is possible without the need to develop an advanced and costly fast reactor technology. The LWBR design relied on a complex movable fuel reactivity control to achieve neutron economy necessary for breeding ratio (BR) higher than unity. Theoretically, Th-{sup 233}U is the only practical combination capable of self-sustainable breeding in the thermal spectrum. Recently, sustainable development with respect to energy and natural resources consumption has become one of the top research priorities. Therefore, Th-based fuel cycle should be reconsidered with a shift in the design objectives, where savings in natural resources becomes important, in addition to non-proliferation and waste issues. In previous studies, we explored the basic possibility of achieving a self-sustainable, with respect to fissile material requirements, Th-{sup 233}U fuel cycle that can be adopted in the current generation of Pressurized Water Reactors. Clearly, the use of movable fuel for reactivity control, as employed in LWBR, is not an option nowadays. However, high conversion ratio can be achieved in LWRs through the use of advanced materials, heterogeneous reactor core structure and careful optimization of fuel composition. Neutronic studies performed on fuel assembly level for the modified PWR assembly designs with BOXER computer code suggested that net breeding of {sup 233}U is possible in principle within a typical PWR operating envelope. As shown previously, considerable core design tradeoffs would be necessary to achieve breeding in PWRs without movable fuel reactivity control. In the new assembly design, most of the initial fissile material is concentrated in relatively small 'seed' regions, which causes a large power peak. Therefore, some reduction in the core power density would be required in order to assure the core safety

  18. Modelling of the local chemistry in stagnant areas in the PWR primary circuit

    International Nuclear Information System (INIS)

    Reid, Rick; Fruzzetti, Keith; Ahluwalia, Al; Summe, Alex; Dame, Cecile; Schmitt, Kyle

    2014-01-01

    MRP-236 demonstrated a correlation between stagnant or low flow conditions and stress corrosion cracking (SCC) of stainless steel components in the PWR primary system. Of the approximately 140 SCC events documented (affecting 15 different components), 83% involved stagnant or low flow conditions that were likely to be associated with chemical environments different from the well mixed bulk coolant. The chemistry in such locations is typically not monitored, and sampling is difficult or impossible. Actions to improve chemistry in regions of low or no coolant flow, such as flushing, cycling of components and imposition of more stringent make up water chemistry controls affect both operational costs and outage schedules. Similarly, design changes to improve flow in affected areas are costly or impracticable. Improving the understanding of the factors controlling chemistry in such areas and development of the capability to predict typical and worst case conditions will allow an informed assessment of procedural actions and/or design changes to improve local chemistry and thereby reduce SCC susceptibility. A project was undertaken to develop a model to predict local chemistry conditions in stagnant locations. The model comprises the iterative application of the EPRI MULTEQ solution chemistry equilibrium code and standard thermodynamic relationships to predict local chemistry conditions considered likely to have been present at the surfaces of components when SCC was initiated. The starting chemistry conditions are based on PWR primary system chemistry from different plant maneuvers (e.g., startup and shutdown conditions). The model was applied to three example components where SCC has occurred in the field. The selected components were: control rod drive mechanism canopy seals; valve drain lines; and reactor vessel o-ring leak-off lines. This paper provides a summary of the model and predicted local chemistry conditions that develop for the three example component as a

  19. The deformation of Zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-08-01

    Simulated PWR fuel rods clad with Zircaloy-4 were tested under convective steam cooling conditions, by pressurising to 0.69-2.07MPa (100-300lb/in 2 ), then ramping at 10 0 C/s to various temperatures in the region 800-955 0 C and holding until either 600 s elapsed or rupture occurred. The length of cladding strained 33% or more was greatest (about 20 times the original diameter) when the initial internal pressure was 1.38+-0.17 PMa (200+-25lb/in 2 ), and the temperature 885 0 C. It is thought that this results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilising the deformation and/or partial superplastic deformation. To avoid adjacent rods in a fuel assembly touching at any temperature, the pressure would have to be less than about 1MPa (145 1b/in 2 ). If the pressure was 1.38MPa (200lb/in 2 ) then the rods would not swell sufficiently to touch if the temperature did not exceed about 840 0 C. (author)

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

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

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

  1. RNL automated ultrasonic inspection of the PISC II PWR inlet nozzle (Plate 3)

    International Nuclear Information System (INIS)

    Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.G.

    1987-01-01

    In June 1984, Risley Nuclear Laboratories (RNL) performed an automated ultrasonic inspection of the Pressurized Water Reactor (PWR) inlet nozzle (plate 3) from the international Programme of Inspection of Steel Components (PISC II) round-robin inspection programme. High-sensitivity pulse-echo detection and predominantly time-of-flight diffraction sizing techniques were employed from the clad inner surface of the nozzle using digital data collection, analysis, and display facilities developed at RNL. RNL detected 30 out of 31 intended weld flaws, achieved one hundred per cent correct acceptance of all acceptable flaws and had a correct rejection frequency on all rejectable flaws of 0.86. The results confirm that well-conceived automated inspection procedures, similar to those used by RNL in this nozzle inspection, could form the basis of a PSI/ISI procedure for reactor pressure vessel nozzle regions. Analysis of the RNL results with regard to the influence of flaw characteristics on inspection performance lends strong support to the general conclusions drawn by the PISC Data Analysis Group. In particular, the most difficult flaws to accurately size were circular smooth and rough flaws. Examination of the RNL results on individual flaws reveals valuable information on the strengths and weaknesses of the adopted procedures and points towards procedural changes that would improve inspection performance. This report describes the procedures adopted by RNL, in the inspection, and reviews the results in the light of definitive flaw information. (author)

  2. Payment Schemes in Conditional Cash Transfer Programs: The Case of 4Ps in the Davao Region, Philippines

    Directory of Open Access Journals (Sweden)

    Ma Cecilia Catubig

    2015-11-01

    Full Text Available This paper evaluates current payment schemes employed by the Pantawid Pamilyang Pilipino Program (4Ps in the Philippines using six assessment criteria: transaction cost, security/risks, speed and timeliness, acceptability, resilience and flexibility. Employing data collected at the regional level, we establish four main findings: (1 all 4Ps payment conduits present trade-offs; (2 a payment approach that uses mainstream financial infrastructure is beneficial if cost, speed and simplicity of the payment system are critical; (3 competition for 4Ps contracts for Payment Service Providers (PSPs has improved the quality of payment services and minimized costs; and (4 the efficiency of the program is greatly influenced by the commitment of the PSP to deliver the cash benefits to the recipients in a timely manner rather than by maximizing conduit branches.

  3. Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

  4. 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

  5. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Nordmann, Francis; Odar, Suat; Rochester, Dewey

    2012-09-01

    Secondary side degradation of steam generators (SG) and Flow Accelerated Corrosion (FAC) in the secondary system have been for a long time important issues in PWR and VVER types of Nuclear Power Plants. With the evolution of the design, the most important issues are progressively moving from secondary side corrosion of Alloy 600 SG tubing, which is being replaced, to a larger variety of risks associated with potential inadequate chemistries. As far as FAC of carbon steel is concerned, the evolution of treatment selection for minimizing corrosion products transport toward the SG, as well as progressive replacement of components in the feedwater train, decreases the risk of dramatic failures which have occurred in the past. After having briefly explained the reason for the past problems encountered in the secondary system of PWR and VVER, this paper evaluates the risk associated with various impurities or contaminants that may be present in the secondary system and how to mitigate them in the most appropriate, efficient, economical and environmental friendly way. The covered species are sodium, calcium, magnesium, chloride, sulfate and sulfur compounds, fluorides, organic compounds, silica, oxygen, lead, ion exchange resins. This paper also proposes the best remedies for mitigating the new issues that may be encountered in operating plants or units under construction. These are mainly: - Selecting a steam water treatment able to minimize the quantity of corrosion products transported toward the SG; - Mitigating the risk of Flow Induced Vibration by a proper control of deposits in sensitive areas; - Minimizing the risk of concentration of impurities in local areas where they may induce corrosion; - Avoiding the presence of abnormal quantities of some species in SG, such as the detrimental presence of lead and ion exchange resin debris or the controversial presence of organic compounds; - Optimizing costs of maintenance activities (SG mechanical, chemical cleaning

  6. Immersed multiple device for the control of the irradiated PWR fuel pins in the reloadable loop in the OSIRIS pond

    International Nuclear Information System (INIS)

    Farny, G.

    1983-01-01

    With respect to the dynamics of the degradation of the PWR fuel in transient, normal and abnormal regions, a new multi-device immersed in the cooling pond of the OSIRIS reactor, is studied. The multiple device is subjected to three examinations: (1) visual studying and video-recording of the appearance of the fuel pins, (2) metrology of the pins, (3) investigation of the induced Foucault currents in the fuel cans. Attention is chiefly paid to the last point; the other ones - being closely related - are only touched on whenever needed. It is concluded that quality control of the fuel pins is possible by means of Foucault currents without applying mechanical constraints and without interfering with the cooling rate. (Auth.)

  7. The role of unconditional cash transfers during a nutritional emergency in Maradi region, Niger: A prospective observational study

    International Nuclear Information System (INIS)

    Fenn, Bridget; Noura, Garba; Sibson, Victoria; Dolan, Carmel; Shoham, Jeremy

    2014-01-01

    Full text: Cash transfers (CTs) are becoming a popular intervention of choice by agencies and NGOs as a complementary or alternative approach to food-based assistance, as part of an emergency response. There is strong evidence that CT programmes lead to an increase in household income and protect household assets from being sold, resulting in an increase in food quantity and improved dietary diversity which in turn are thought to protect children from malnutrition. However, the evidence for an impact of CTs on undernutrition is mixed and inconclusive. Despite this, CTs are increasingly being used in emergency responses with an objective of preventing acute malnutrition. The main objective was to assess the effect of an unconditional CT implemented as part of an emergency response to food insecurity during a declared state of emergency in Aguie district, Maradi, Niger. This was a prospective observational study involving 6 consecutive months of data collection starting pre-intervention in April 2012 (baseline), on the same cohort of ‘poor’ and ‘very poor’ households, with a non-acutely malnourished child 6-36 months, enrolled by Save the Children in an unconditional CT programme (n = 412). Analyses using pre-post intervention data were carried out to assess changes in the potential mediating factors within the causal pathway between CT and acute malnutrition over time and to estimate risk factors associated with acute malnutrition. The study showed that the living standards of ‘poor’ and ‘very poor’ households improved; indicated by reduction in poverty (improvement in household expenditures, incomes, employment, asset protection, wealth rank and access to social networks) and improvement in household food security (reduced household hunger and greater household and child dietary diversity). Child anthropometric outcomes (weight-for-height and MUAC) significantly improved (p<0.001), despite a decline in child health and women’s well-being and

  8. The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

    Science.gov (United States)

    Cockeram, B. V.; Kammenzind, B. F.

    2018-02-01

    Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

  9. Evaluation of a knowledge transfer scheme to improve policy making and practices in health promotion and disease prevention setting in French regions: a realist study protocol.

    Science.gov (United States)

    Cambon, Linda; Petit, Audrey; Ridde, Valery; Dagenais, Christian; Porcherie, Marion; Pommier, Jeanine; Ferron, Chrisine; Minary, Laetitia; Alla, François

    2017-06-29

    Evidence-based decision-making and practice are pivotal in public health. However, barriers do persist and they relate to evidence properties, organisations and contexts. To address these major knowledge transfer (KT) issues, we need to rethink how knowledge is produced and used, to enhance our understanding of decision-making processes, logics and mechanisms and to examine the ability of public health services to integrate research findings into their decisions and operations. This article presents a realist evaluation protocol to assess a KT scheme in prevention policy and practice at local level in France. This study is a comparative multiple case study, using a realist approach, to assess a KT scheme in regional health agencies (ARS) and regional non-profit organisations for health education and promotion (IREPS), by analysing the configurations contexts/mechanisms/outcomes of it. The KT scheme assessed is designed for the use of six reviews of systematic reviews concerning the following themes: nutrition, alcohol, tobacco smoking, physical activity, emotional and sexual life and psychosocial skills. It combines the following activities: supporting the access to and the adaptation of scientific and usable evidences; strengthening professionals' skills to analyse, adopt and use the evidences in the course of their practices and their decision-making process; facilitating the use of evidence in the organisations and processes. RAMESE II reporting standards for realist evaluations was used. The aims of this study are to experiment and characterise the factors related to the scheme's ability to enable public health stakeholders to address the challenges of KT and to integrate scientific knowledge into policy and practice. We will use the realist approach in order to document the parameters of successful KT strategies in the specific contexts of preventive health services in France, while seeking to determine the transferability of such strategies.

  10. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  11. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat-transfer surfaces

    International Nuclear Information System (INIS)

    Burgmayer, P.; Crovetto, R.; Turner, C.; Klimas, S.J.

    1999-07-01

    The effectiveness of 3 different dispersants-a polyphosphonic acid (PIPPA), a polymethacrylic acid (PMA), and a hydroxyethylidene methacrylic acid (HEME)-at controlling magnetite deposition was examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products of a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using an 59 Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any effect on deposition under single-phase forced-convective flow. (author)

  12. Effectiveness of selected dispersants on magnetite deposition at simulated PWR heat transfer surfaces

    International Nuclear Information System (INIS)

    Burgmayer, P.; Crovetto, R.; Turner, C.; Klimas, S.

    1998-01-01

    The effectiveness of three different dispersants - a polyphosphonic acid (PIPPA); a polymethacrylic acid (PMA); and a hydroxyethylidene methacrylic acid (HEME) - at controlling magnetite deposition has been examined under steam generator operating conditions. Tests in a cycling research model boiler showed that the dispersants resulted in corrosion products with a smaller average size and a bimodal size distribution. At a concentration in the boiler of 10 mg/kg, density weight deposit on heated probes was reduced 4-, 3-, and 2-fold for PMA, PIPPA, and HEME, respectively. PIPPA was the most effective at increasing iron transport out of the boiler. In deposition loop tests using a 59-Fe radiotracer, only PIPPA and HEME were effective at reducing the particle deposition rate under flow-boiling conditions. None of the dispersants had any impact on deposition under single-phase forced-convective flow. (author)

  13. A new model for simulation of pressurizers in PWR type reactors

    International Nuclear Information System (INIS)

    Madeira, A.A.; Oliveira Barroso, A.C. de

    1981-01-01

    The pressurizer is treated as three-homogeneous region thermodynamic system, whith movable boundaries, all regions considered under the same pressure. In Normal operation, the two botton regions are occupied by water (liquid), and steam occupies the top region. Normal and spray induced condensation processes, evaporation and heat transfer cross the steam-water interface are analysed. The liquid region at the very botton of the pressurizer is treated in a simplyfied manner in order to retain the computacional advantages of the two-region models. (Author) [pt

  14. Sizewell B: consent application for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1981-02-01

    The Central Electricity Generating Board has applied to the Secretary of State for Energy for consent and for other necessary permissions to construct a nuclear power station of about 1200 MW output capacity based on the pressurised water reactor (PWR) system on the Board's existing site at Sizewell (near Leiston) in Suffolk to be known as Sizewell B. Application has also been made to the Health and Safety Executive to extend the existing nuclear site licence to permit the use of the site for a pressurised water reactor. The Secretary of State for Energy has already stated that a Public Inquiry will be held into the application and this is expected to take place in 1982. The Board is making these applications now to give ample time for public discussion and consultation. Construction of the station could not begin until the outcome of the Public Inquiry is known and the necessary consents, nuclear licence and clearances have been given. The text of the application is presented. Some background information is given. (author)

  15. Effect of component aging on PWR control rod drive systems

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

  16. Plutonium thermal utilization in PWR in Mihama No. 1 plant

    International Nuclear Information System (INIS)

    Yokote, Mitsuhiro; Kondo, Yoshiaki; Shimada, Shouichirou; Abeta, Sadaaki.

    1992-01-01

    On December 20, 1991, the use of four MOX fuels charged in Mihama No. 1 plant for three cycles ended, which is the verification project with small number of specimens on the plutonium thermal utilization in PWRs in Japan. There was not any symptom of showing abnormality in the safety of the core and the soundness of the fuel during the use. In this report, the verification project and the results are explained. In spent fuel, reusable fission substances such as Pu-239 and Pu-241 produced from U-235 and U-238 are contained. By recycling and effectively utilizing them, resources are protected and the effect to environment is reduced, the energy security in Japan with poor resources can be heightened, and waste management becomes proper. The course of the plutonium thermal utilization in PWR project in Mihama No. 1 plant, the design of MOX fuel and the core, the manufacture of MOX fuel in USA and its transport to Japan, the preservation, practical use and operation management of MOX fuel, the charging of MOX fuel in Mihama No. 1 plant and the use, and the plan of the plutonium thermal utilization in PWRs for hereafter are reported. (K.I.)

  17. Fluid-structure interactions in PWR vessels during blowdown

    International Nuclear Information System (INIS)

    Schumann, U.; Enderle, G.; Katz, F.; Ludwig, A.; Moesinger, H.; Schlechtendahl, E.G.

    1979-01-01

    For analysis of blowdown loadings and dynamic response of PWR vessel internals several computer codes have been developed at Karlsruhe. The goal is to provide advanced codes which permit a 'best estimate' analysis of the deformations and stresses of the internal structures, in particular the core barrel, such that the safety margins can be evaluated. The stresses reach their maxima during the initial subcooled period of the blowdown in which two-phase phenomena are important in the blowdown pipe only. In this period, the computed results with and without fluid-structural interactions show that the coupling between the water in the downcomer and the rather thin elastic core barrel is of dominant importance. Without coupling the core barrel oscillates with much higher frequencies than with coupling. The amplitudes and stresses are about twice as large initially. Later, the decoupled analysis can result in a meaningless overestimation of the structural response. By comparison of computations for incompressible and for compressible fluid with and without coupling we have found that a correct treatment of the fluid-structure coupling is more important than the description of pressure waves. (orig.)

  18. 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.

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    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.

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

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    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

  2. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1989-01-01

    The report presents experimental results of a small lower plenum break test of SB-PV-01 conducted at the large-Scale Test Facility (LSTF) of the Rig-of-Safety Assessment (ROSA)-IV program. This test simulates a loss-of-coolant accident (LOCA) caused by instrument tubes break (break area corresponds to 0.5% of the cold leg flow area) in a Westinghouse-type pressurized water reactor (PWR) assuming both manual actuation for all of the high pressure injection (HPI) systems and failure of the auxiliary feedwater systems. The report clarifies long-term system responses, especially the core cooling conditions related to the primary mass inventory. Also it clarifies break location effects on small break LOCA phenomena by comparing other five similar LOCA tests with break locations at cold leg, hot leg, upper head, pressurizer top (TMI-type) and SG U-tubes. It is coucluded that the lower plenum break is the severest on core heatup due to the highest break flow rate and the least primary mass recovery after the ECCS among the six tests. (author)

  3. Source term aspects associated with future PWR containment systems

    International Nuclear Information System (INIS)

    Kuczera, B.; Kebler, G.; Ehrhardt, J.; Scholtyssek, W.

    1994-01-01

    The overall objective of reactor safety is to protect the population against dangerous releases of radioactive materials from nuclear power plants. In context with a reinforcement of the defense-in-depth strategy the common safety requirements on future nuclear power plants converge in the objective that these plants should be so safe that even in case of a severe accident there will be no need of off-site emergency actions such as an evacuation or resettlement of the population from the vicinity of a nuclear power plant. It is shown by the example of a future 1400 MWe pressurized water reactor (PWR) plant that this goal can be attained in principle by providing a double containment with the annulus vented via an appropriate emergency standby filter. Within the framework of severe accident consequence mitigation a set of parameters for accident conditions and emergency filter efficiencies is elaborated under which the German lower levels of intervention for evacuation are not attained. (author). 10 refs., 3 tabs., 5 figs

  4. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  5. Aerosols behavior inside a PWR during an accident

    International Nuclear Information System (INIS)

    Hervouet, C.

    1983-01-01

    During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems) [fr

  6. Water chemistry control of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Hino, Yuichi; Makino, Ichiro; Yamauchi, Sumio; Fukuda, Fumihito.

    1992-01-01

    In PWR power plants, the primary system taking heat out of nuclear reactors and the secondary system generating steam and driving turbines are completely separated by steam generators, accordingly, by mutually independent water treatment, both systems are to be maintained in the optimal conditions. Namely, primary system is the closed water circulation circuit of simple liquid phase though under high temperature, high pressure condition, therefore, water shows the stable physical and chemical properties, and the minute water treatment for restraining the corrosion of structural materials and reducing radioactivity can be done. Secondary system is similar to the condensate and feedwater system of thermal power plants, and is the circuit for liquid-vapor two-phase transformation, but due to the local concentration of impurities by evaporation, the strict requirement is set for secondary water quality. However, secondary system can be treated in the state without radioactivity, and this is a great merit. The outline, basic concept and execution of primary water quality control, and the outline, concept, control criteria, facilities and execution of secondary water quality control are reported. (K.I.)

  7. Qualification tests for PWR control element drive mechanism

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Yong; Jin, Choon Eon; Choi Suhn [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new.

  8. Improvements of nuclear fuel management in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1978-07-01

    The severe variations to which the different elements contributing to the determination of the fuel cycle cost are subjected have led to a reopening of the problem of ''optimization'' of nuclear fuel management. The increase in costs of uranium ore, isotope separation work units (swu), reprocessing, the political implications of proliferation associated with the employment of reprocessing operations have been at the origin of a reassessment of present-day management. It therefore appeared to be appropriate to study variants with respect to a reference mode represented by the management of the PWR 900 MWe systems, without burnable poison in the cycle at equilibrium (Case 3 of Table 1). In order to obtain a complete view of impacts of such modifications, computations were carried out as far as the appraisal of the cycle cost and with reprocessing. There has likewise been added to this the estimate of the gain anticipated from certain improvements in the neutron balance contributed at the level of the lattice

  9. Integrated training support system for PWR operator training simulator

    International Nuclear Information System (INIS)

    Sakaguchi, Junichi; Komatsu, Yasuki

    1999-01-01

    The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

  10. Tendency of nuclear pumps for PWR primary system

    International Nuclear Information System (INIS)

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  11. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    Lambert, M.; Salvatores, St.; Ferrier, A.; Pelet, J.; Nicaise, N.; Pouliquen, J.Y.; Foret, F.; Chauliac, C.; Johner, J.; Cohen, Ch.

    2003-01-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  12. Robots in P.W.R. nuclear powerplants

    International Nuclear Information System (INIS)

    Dubourg, M.

    1987-01-01

    The satisfactory operation of 37 900-MWe PWR powerplants in France, Belgium and South-Africa and the start-up of 1300 MWe powerplants allowed the development of a wide range of automatic units and robots for the periodic maintenance of nuclear plants, reducing the risk of ionizing radiation for the personnel. A large number of automated tools have been built. Among them: - inspection and maintenance systems for the tube bundle of steam generators, - robotized arms ROTETA and ROMEO for the heavy maintenance and delicate operations such as tube extraction or shot peening of tubes to improve their resistance to corrosion; - the versatile manipulator T.A.M. with electrically controlled articulations. The development of functionally versatile tools and robots and the integration of new technologies such as 3-D vision allowed the construction of the self-guided vehicle FRASTAR capable of moving within a nuclear building and in a cluttered environment. This vehicle includes means for avoiding isolated obstacles and can move on stairs [fr

  13. Control rod effects with plutonium recycle in a PWR

    International Nuclear Information System (INIS)

    Nash, G.; Muehl, G.J.; Gibson, I.H.

    1979-03-01

    A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

  14. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  15. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.

    2013-01-01

    Highlights: ► Numerical results reveal that the HS method is reliable. ► The great advantage of HS is significant gain in computational cost. ► On the average, the final band width of search fitness values is narrow. ► Our experiments show that the search approaches the optimal value fast. - Abstract: In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrangement of fuel assemblies in PWR cores. This code consists of two sections including Harmony Search algorithm and Nodal Expansion modules using fourth degree flux expansion which solves two dimensional-multi group diffusion equations with one node per fuel assembly. Two optimization test problems are investigated to demonstrate the HS algorithm capability in converging to near optimal loading pattern in the fuel management field and other subjects. Results, convergence rate and reliability of the method are quite promising and show the HS algorithm performs very well and is comparable to other competitive algorithms such as Genetic Algorithm and Particle Swarm Intelligence. Furthermore, implementation of nodal expansion technique along HS causes considerable reduction of computational time to process and analysis optimization in the core fuel management problems

  16. Development of the business area construction and energy of EnergieRegion Nuernberg. Transfer from project management to a regional network; Entwicklung des Geschaeftsfeldes Bau und Energie der EnergieRegion Nuernberg. Umsetzung von Projektmanagement in einem regionalen Netzwerk

    Energy Technology Data Exchange (ETDEWEB)

    Seiverth, A.

    2006-07-01

    The association EnergieRegion Nuernberg is a regional authority network, which is employed with the promotion of sustainable handling of the factor energy in the region Nuernberg and with the proliferation of this region as internationally recognized location for energy engineering, energy industry and energy science. The intention is to use the important industrial, service-oriented and scientific potential optimally. For this reason a functional co-ordination and communication platform had to be created for the cross-linking of the appropriate participants from economics, research and public administration. Therefore, the author of the contribution under consideration accompanies the development process of the business field construction and energy of this association in the background of the current trends in the construction and energy sector in the region Nuernberg. Under this aspect, the author reports on the following aspects: (a) Success factors of the project management in a regional network; (b) Operationalisation of the success of the project by means of a model; (c) Analysis of the different aspects of energetic measures; (d) Determination of chances and risks of the range building and energy in the region Nuernberg; (e) Comparison of the success of the model projects with the model for the determination of project success; (f) Determination of strengths and weaknesses of the project management in the business field construction and energy of the energy region Nuernberg.

  17. Application of modeling to local chemistry in PWR steam generators

    International Nuclear Information System (INIS)

    Fauchon, C.; Millett, P.J.; Ollar, P.

    1998-01-01

    Localized corrosion of the SG tubes and other components is due to the presence of an aggressive environment in local crevices and occluded regions. In crevices and on vertical and horizontal tube surfaces, corrosion products and particulate matter can accumulate in the form of porous deposits. The SG water contains impurities at extremely low levels (ppb). Low levels of non-volatile impurities, however, can be efficiently concentrated in crevices and sludge piles by a thermal hydraulic mechanism. The temperature gradient across the SG tube coupled with local flow starvation, produces local boiling in the sludge and crevices. Since mass transfer processes are inhibited in these geometries, the residual liquid becomes enriched in many of the species present in the SG water. The resulting concentrated solutions have been shown to be aggressive and can corrode the SG materials. This corrosion may occur under various conditions which result in different types of attack such as pitting, stress corrosion cracking, wastage and denting. A major goal of EPRI's research program has been the development of models of the concentration process and the resulting chemistry. An improved understanding should eventually allow utilities to reduce or eliminate the corrosion by the appropriate manipulation of the steam generator water chemistry and or crevice conditions. The application of these models to experimental data obtained for prototypical SG tube support crevices is described in this paper. The models adequately describe the key features of the experimental data allowing extrapolations to be made to plant conditions. (author)

  18. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  19. PWR-PSMS benchmarking results using thermocouple data from the summer-1 plant

    International Nuclear Information System (INIS)

    Peng, C.M.; Ipakchi, A.; Kim, J.H.

    1986-01-01

    In large pressurized water reactor (PWR) power plants, estimating the in-core power distribution from off-line predictions is based on data from global measurements with conservative assumptions. The off-line predictions are too independent of the actual process to reflect the true state of the reactor. The on-line core monitoring systems tend to balance between measurements and theoretical calculations, better utilizing information coming from measurements. The hybrid system, which incorporates measurements in predictions along with frequent model adaptations, will closely track the actual operating state of the plant. Since the detailed core flux mapping is performed with large time intervals for those PWRs without fixed in-core detectors, the on-line signals from thermocouples located at the top of selected fuel assemblies offer an alternative means of monitoring. The in-core thermocouples give a good indication of the average coolant temperature at the outlet of the instrumented assemblies and potentially can provide continuous information of the radial power distribution between flux maps. The PWR Power Shape Monitoring System (PWR-PSMS) has implemented this on-line monitoring feature based on thermocouple readings to evaluate the core performance and to improve core monitoring. The purpose of this paper is to present the benchmark results of PWR-PSMS using thermocouple data from the Summer-1 plant of a Westinghouse PWR

  20. Modelling of CRUD growth phenomena on PWR fuel rods under nucleate boiling conditions

    International Nuclear Information System (INIS)

    Ferrer, A.; Dacquait, F.; Gall, B.; Ranchoux, G.; Riot, G.

    2012-09-01

    PWR primary circuit materials undergo general corrosion leading to a release of metallic element release and subsequent process of particle deposition and ion precipitation on the primary circuit surfaces. The species accumulated on fuel rods are activated by neutron flux. Consequently, crud erosion and dissolution induce primary coolant contamination. In French PWRs, 58 Co volume activity is generally low and almost constant (< 30 MBq.m -3 ) throughout an ordinary operating cycle. In some specific cases, a significant increase in volume activity is observed after the middle of a cycle (100-1000 MBq.m -3 for 58 Co) when conditions for nucleate boiling are locally reached in certain fuel assemblies. Indeed, it is well known that nucleate boiling intensifies the deposition process. The thickness of the crud layer can reach some micrometers in non-boiling areas, whereas it can reach 100 micrometers in boiling areas. Crud growth in boiling conditions can be related to three phenomena: bubble growth induces deposition process (called boiling deposition), bubbles induce concentration increase at crud-coolant interface (called enrichment and modelled by the enrichment factor, the ratio between the wall concentration and the bulk concentration) and vaporisation induces concentration increase inside the crud. A literature review on the modelling of these phenomena and on the crud structure in nucleate boiling conditions has been carried out. The OSCAR [1] calculation code developed by the CEA to predict surface and volume activities in a single phase PWR primary circuit was chosen as a basis for present study. Ability to describe local nucleate boiling conditions was added to this code leading to realistic modelling of subsequent volume activity increase. In this article, we present the results obtained using a modified version of the OSCAR PC V1.2 calculation code including: - A double phase thermal-hydraulic module, - A model of boiling crud growth, able to calculate

  1. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    Science.gov (United States)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  2. Transfer Zymography.

    Science.gov (United States)

    Pan, Daniel; Wilson, Karl A; Tan-Wilson, Anna

    2017-01-01

    The technique described here, transfer zymography, was developed to overcome two limitations of conventional zymography. When proteolytic enzymes are resolved by nonreducing SDS-PAGE into a polyacrylamide gel with copolymerized protein substrate, the presence of the protein substrate can result in anomalous, often slower, migration of the protease and an estimated mass higher than its actual mass. A further drawback is that the presence of a high background of substrate protein interferes with proteomic analysis of the protease band by excision, tryptic digestion, and LC-MS/MS analysis. In transfer zymography, the proteolytic enzymes are resolved by conventional nonreducing SDS-PAGE, without protein substrate in the gel. The proteins in the resolving gel are then electrophoretically transferred to a receiving gel that contains the protein substrate, by a process similar to western blotting. The receiving gel is then processed in a manner similar to conventional zymography. SDS is removed by Triton X-100 and incubated in conditions suitable for the proteolytic activity. After protein staining, followed by destaining, bands representing regions with active protease are visualized as clear bands in a darkly stained background. For proteomic analysis, electrophoresis is carried out simultaneously on a second resolving gel, and the bands corresponding to the clear regions in the receiving gel after zymogram development are excised for proteomic analysis.

  3. Hydrogeological boundary settings in SR 97. Uncertainties in regional boundary settings and transfer of boundary conditions to site-scale models

    International Nuclear Information System (INIS)

    Follin, S.

    1999-06-01

    The SR 97 project presents a performance assessment (PA) of the overall safety of a hypothetical deep repository at three sites in Sweden arbitrarily named Aberg, Beberg and Ceberg. One component of this PA assesses the uncertainties in the hydrogeological modelling. This study focuses on uncertainties in boundary settings (size of model domain and boundary conditions) in the regional and site-scale hydrogeological modelling of the three sites used to simulating the possible transport of radionuclides from the emplacement waste packages through the host rock to the accessible environment. Model uncertainties associated with, for instance, parameter heterogeneity and structural interpretations are addressed in other studies. This study concludes that the regional modelling of the SR 97 project addresses uncertainties in the choice of boundary conditions and size of model domain differently at each site, although the overall handling is acceptable and in accordance with common modelling practice. For example, the treatment of uncertainties with regard to the ongoing post-glacial flushing of the Baltic Shield is creditably addressed although not exhaustive from a modelling point of view. A significant contribution of the performed modelling is the study of nested numerical models, i.e., the numerical interplay between regional and site-scale numerical models. In the site-scale modelling great efforts are made to address problems associated with (i) the telescopic mesh refinement (TMR) technique with regard to the stochastic continuum approach, and (ii) the transfer of boundary conditions between variable-density flow systems and flow systems that are constrained to treat uniform density flow. This study concludes that the efforts made to handle these problems are acceptable with regards to the objectives of the SR 97 project

  4. 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.

  5. Hydraulic test for non-instrumented capsule of advanced PWR fuel pellet

    International Nuclear Information System (INIS)

    Jun, Hyung Gil; Yoon, Y. J.; Chun, S. Y.; Kim, D. H.; Lee, C. B.; Ryu, J.

    2001-04-01

    This report presents the results of pressure drop test, vibration test and endurance test for Non-instrumented Capsule of Advanced PWR Fuel Pellet which were designed fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate across the Non-instrumented Capsule of Advanced PWR Fuel Pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the Non-instrumented Capsule of Advanced PWR Fuel Pellet ranges from 13.0 to 32.3 Hz. RMS(Root Mean Square) displacement for the fuel rig is less than 11.6 μm, and the maximum displacement is less than 30.5 μm. The endurance test was carried out for 103 days and 17 hours

  6. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  7. Evaluation and categorization of secondary system layup and cleanup practices for PWR plants

    International Nuclear Information System (INIS)

    Cleary, W.F.

    1982-12-01

    The EPRI Program S113-1, Evaluation of Secondary System Layup and Cleanup Proctices was established to study ways to minimize the transport of corrosion products into the secondary side PWR steam generators that occurs during plant startups following extended outages. As part of the EPRI Program, Task 200 objective was to identify and categorize the layup and cleanup practices now in use or proposed by utilities for PWR plants. The task study consisted of gathering information by conducting site visits to fourteen representative PWR plants in the USA, Europe and Japan, by conducting a search of the open literature, reviews of related EPRI Programs, and by evaluating the practices in terms of their potential effectiveness. The results show that about 30% of the plants attempt routine layup of secondary systems during outages and about 60% perform some form of system cleanup during the return to power following extended outages

  8. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    Energy Technology Data Exchange (ETDEWEB)

    Putney, J.M.; Preece, R.J. [National Power, Leatherhead (GB). Technology and Environment Centre

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  9. Assessment of PWR Steam Generator modelling in RELAP5/MOD2

    International Nuclear Information System (INIS)

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3

  10. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  11. Laser based maintenance technology for PWR power plants

    International Nuclear Information System (INIS)

    Itaru Chida; Masaki Yoda; Naruhiko Mukai; Yuji Sano; Makoto Ochiai; Takahiro Miura; Ryoichi Saeki

    2005-01-01

    Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance technologies and already applied them to several existing nuclear power plants. Recently, we have developed the maintenance system for the inner surface of bottom-mounted instruments (BMI) of PWR plants. This system performs nondestructive testing (NDT) and preventive maintenance against SCC by using YAG lasers. Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed by using a compact probe with a multi-mode optical fiber and an interferometer. This system is used for both detection and depth measurement of surface-breaking cracks. It is confirmed through laboratory studies that the developed system successfully detected and sized micro slits of around 1.0 mm depth on weld metal and heat-affected zone (HAZ). SCCs produced by chemical method were also tested by the system. For the preventive maintenance treatment, laser-peening (LP) technology was developed and already applied to several reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP (FLP) system as a preventive maintenance against SCC. For checking the effect of FLP, we carried out FLP experiments on the inner surface of a small tube-shaped Alloy 600 by using this system. After FLP, residual stress was measured by X-ray method for radial and axial directions on the inner surface of the tube, and effectiveness of stress improvement was proved. Based on these experiences, LUT and FLP were applied to Ikata unit-1 of Shikoku Electric Power Company Inc. and successfully treated the inner surface of BMIs. (authors)

  12. Radiation detectors for the control of PWR nuclear boilers

    International Nuclear Information System (INIS)

    Duchene, J.

    1977-01-01

    The neutronic control in French PWR is effected by: 2 channels of measurement of intermediate power using γ'-compensated boron-coated ionization chambers 4 channels of measurement of high power with 'long' boron chambers also used in axial off-set measurement. A movable in-core measuring system is used for the fuel management and the power distribution monitoring. The instrumentation of start-up and intermediate power is conventional; the chambers of the axial off-set measurement and the in-core system are special for this type of power plant, they are discussed in details. The essential properties of the various types of detector, their major advantages or drawbacks, their comparative adaptation to the functions to be performed in the plant are summarized in a table. The 'long chambers' (on use in Fessenheim I and II, and soon in Bugey II) are boron coated current ionization chambers, without γ compensation, intended for power measurement. In-core measurements first involved activation methods - movable wires giving flux profiles, -or activable nuts (the Aeroball System at Trino Vercellese, Chooz...). In on-line neutron detectors, used at fixed positions, the electric signal is generated from: ionization the gas filling fission ionization chambers and γ ionization chambers; direct collection of the charged particles emitted from the convertor element in self-powered neutron detectors (rhodium, silver or vanadium) or self-powered γ detectors (cobalt); or thermoelectric effect in neutron and γ thermometers. The in-core measurement unit developped by Framatome is a movable miniaturized fission chamber system (at Tihange), every French exported power plant being now equipped with it [fr

  13. Draining water loop seals in a PWR plant

    International Nuclear Information System (INIS)

    Bhavnani, D.; Flaherty, J.; Coward, B.; Gorga, J.

    1994-01-01

    Pressurized Water Reactor (PWR) power plants include safety valves (SVs) on the pressurizer to provide over-pressure protection for the primary coolant system. In addition, power operated relief valves (PORV) are also included to allow pressure control. These valves are located in piping connecting the pressurizer to the pressurizer relief tank (PRT). In some plants, the SV inlet piping is oriented to specifically form a water loop seal adjacent to the valves. Steam from the pressurizer enters the piping and condenses to form a water seal against the valve. The water seal provides protection for the valve's internals and creates a better valve seal. Additionally, the PORV inlet piping may also be oriented to form a water loop seal similar to that for SVs. The SVs and PORVs are normally closed. During an over-pressure transient, the valves open and the water seals discharge through the valves and downstream piping to the PRT. This sudden discharge of the water slug through piping normally containing low pressure steam or air can cause significant unbalanced hydrodynamic forces on the piping and cause piping or pipe support damage. Utilities must demonstrate that these forces will not cause sufficient damage such that the piping will no longer function as designed. This paper describes a method for reducing these unbalanced forces by installing drains that allow the condensed loop seal water to flow back into the pressurizer. This approach, which is a passive modification in which no active components are added, reduces the mass of water available for acceleration through the valve and piping, significantly decreasing the hydrodynamic forces. Another important consideration is that the modification has little or no effect on plant operation and maintenance. Thermal hydraulic analyses are performed to estimate the hydrodynamic forces and time history finite element stress analyses are performed to calculate pipe stress and pipe support loads

  14. An EPRI perspective and overview of PWR primary chemistry optimization

    International Nuclear Information System (INIS)

    Perkins, David; Haas, Carey; Kucuk, Aylin; Reid, Rick

    2009-01-01

    Initiatives are underway to optimize primary water chemistry to promote long term equipment reliability, dose and fuel deposit management, and maintenance of system and core integrity. These initiatives include increased primary system pH(t), zinc injection, and optimization of primary system hydrogen concentration. The concurrent demands of higher core power densities and longer operating cycles make implementation and evaluation of such chemistry changes increasingly challenging to plant chemists and operators. One of the most significant changes has been the injection of zinc. The primary reason for zinc injection is dose reduction as part of an overall dose management program. Since initial implementation in 1994, zinc injection has been successfully initiated at more than 60 Pressurized Water Reactors (PWRs) worldwide. This equates to approximately 23% of the operating PWR's. Current projections show that greater than 25% of the fleet will be injecting zinc by the end of 2010. The EPRI Materials Reliability Program (MRP), Fuel Reliability Program (FRP) and Chemistry program have ongoing research related to zinc injection and elevated hydrogen to support industry efforts in dose reduction, mitigation of PWSCC in nickel-based alloys and improved fuel reliability. Fuel performance, effects on plant materials and safety implications must be considered prior to modification of primary system chemistry controls. Evaluation of these effects typically requires additional research, which may include fuel performance monitoring and post-shutdown fuel surveillances to understand and evaluate the impact of changes on system and fuel performance. The poster describes ongoing industry experience(s) and research work in the EPRI Chemistry, FRP, and MRP areas related to ongoing primary chemistry programmatic changes. (author)

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

    International Nuclear Information System (INIS)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R.

    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

  16. The reliability data acquisition system in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Lienart, P.

    1984-01-01

    In April 1978, Electricite de France put a reliability data acquisition system (SRDF) into operation at its two nuclear power plant sites: Fessenheim and Bugey. In the light of the experience acquired and the advantages offered by such a data bank, this system has been progressively extended since 1982 to cover the entire PWR network. The SRDF was originally designed for the follow-up of 4000 items of equipment per pair of units. However, the various difficulties encountered in gathering data made it necessary - in order to safeguard the quality of the information - to reduce this number initially to 800 major mechanical or electromechanical items of equipment designed to ensure the safety or availability of the units. Subsequently, an increase to 1100 was possible. The SRDF consists of a centralized information bank linked by telephone to the various nuclear sites. The software enables the data-acquisition cards to be introduced, modified or deleted. Any user can gain access to the bank by simply making queries in real time. The quality of the acquisition and processing of the data depend on a list of equipment confined to essential operational systems and on a card design combining, as far as possible, the precision and accessibility of the data. A method of logical failure analysis has also been devised, its main purposes being to provide the following: (1) aid to card instruction; (2) an easier way of checking the uniformity of information concerning a failure; and (3) compatibility between the instructions and analysis of data, thereby facilitating development of the data-processing program. (author)

  17. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  18. Contribution to the study of the conversion PWR type reactors to the thorium cycle

    International Nuclear Information System (INIS)

    Martins Filho, J.R.

    1980-01-01

    The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author) [pt

  19. Scope and procedures of fuel management for PWR nuclear power plant

    International Nuclear Information System (INIS)

    Yao Zenghua

    1997-01-01

    The fuel management scope of PWR nuclear power plant includes nuclear fuel purchase and spent fuel disposal, ex-core fuel management, in-core fuel management, core management and fuel assembly behavior follow up. A suit of complete and efficient fuel management procedures have to be created to ensure the quality and efficiency of fuel management work. The hierarchy of fuel management procedure is divided into four levels: main procedure, administration procedure, implement procedure and technic procedure. A brief introduction to the fuel management scope and procedures of PWR nuclear power plant are given

  20. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  1. Assessment of options for the treatment of Sizewell PWR liquid effluent

    International Nuclear Information System (INIS)

    Hornby, J.; Allam, J.; Knibbs, R.H.

    1992-01-01

    This report describes the origins of PWR liquid waste streams, their composition and rates of arising. Data has been collected from operational PWRs and estimates obtained for Sizewell B PWR liquid waste streams. Current liquid waste treatment practices are reviewed and assessments made of established and novel treatment techniques which could be applicable to Sizewell B. A short list of treatment options is given and recommendations are made relating to established treatment technologies suitable for Sizewell B and also to development work on more novel treatments which could lead to a reduction in waste disposal volumes. (author)

  2. Effect of TOC [total organic carbon] on a PWR secondary cooling water system

    International Nuclear Information System (INIS)

    Gau, J.Y.; Oung, J.C.; Wang, T.Y.

    1989-01-01

    Increasing the amount of total organic carbon (TOC) during the wet layup of the steam generator was a problem in PWR nuclear power plant in Taiwan. The results of surveys of TOC in PWR secondary cooling water systems had shown that the impurity of hydrazine and the bacteria were the main reasons that increase TOC. These do not have a corrosion effect on Inconel 600 and carbon steel when the secondary cooling water containing the TOC is below 200 ppb. But the anaerobic bacteria from the steam generator in wet layup will increase corrosion rate of carbon steel and crevice corrosion of Inconel 600. (author)

  3. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  4. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  5. On-line analysis of ETA and organic acids in secondary systems of PWR plants

    International Nuclear Information System (INIS)

    Kurashina, Masahiko; Uzawa, Hideo; Utagawa, Koya; Takaku, Hiroshi

    2005-01-01

    To reduce the iron concentration in the secondary water of plants with pressurized water reactors (PWRs), ethanolamine (ETA) is used as an alkalizing agent in the secondary cycle. An on-line ion chromatography (IC) monitoring system for monitoring concentrations of ETA and anions of organic acids was developed, its performance was evaluated, and verification tests were conducted at an actual PWR plant. It was demonstrated that the concentration of both ETA and anions of organic acids may be successfully monitored by IC in PWR secondary cycle streams alkalized by ETA. (orig.)

  6. Prevention and mitigation of steam-generator water-hammer events in PWR plants

    International Nuclear Information System (INIS)

    Han, J.T.; Anderson, N.

    1982-11-01

    Water hammer in nuclear power plants is an unresolved safety issue under study at the NRC (USI A-1). One of the identified safety concerns is steam generator water hammer (SGWH) in pressurized-water reactor (PWR) plants. This report presents a summary of: (1) the causes of SGWH; (2) various fixes employed to prevent or mitigate SGWH; and (3) the nature and status of modifications that have been made at each operating PWR plant. The NRC staff considers that the issue of SGWH in top feedring designs has been technically resolved. This report does not address technical findings relevant to water hammer in preheat type steam generators. 10 figures, 2 tables

  7. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

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

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  10. Analysis of the alternatives for the chemical treatment of the secondary circuit of PWR power plants

    International Nuclear Information System (INIS)

    Lopes, J.P.G.; Silva Neto, A.J. da; Braganca Junior, A.; Dominguez, D.

    1990-01-01

    The operational experiences within PWR power plants shows that the major problems which affect the plant availability occurs in the secondary side, mainly in the steam generators and condenser. The aim of this report is to perform an evaluation of the main chemical treatment processes, which are applied to the secondary side of PWR power plants in order to reduce the corrosion problems to which are exposed the system equipment, minimizing in this way the shut down and maintenance cost for repairs and replacement of damaged components. (author)

  11. Evaluation of PWR steam generator water hammer. Final technical report, June 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    Block, J.A.; Crowley, C.J.; Rothe, P.H.; Wallis, G.B.; Young, L.R.

    1977-05-01

    An investigation of waterhammer in the main feedwater piping of PWR steam generators due to water slugs formed in the steam generator feedring is reported. The relevant evidence from PWR operation and testing is compiled and summarized. The state-of-the-art of analysis of related phenomena is reviewed. Original exploratory modeling experiments at 1 / 10 and 1 / 4 scale are reported. Bounding analyses of the behavior are performed and several key phenomena have been identified for the first time. Recommendations to the Nuclear Regulatory Commission are made

  12. Categorization of PWR accident sequences and guidelines for fault trees: seismic initiators

    International Nuclear Information System (INIS)

    Kimura, C.Y.

    1984-09-01

    This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to determine the applicability of the dominant accident sequences of Zion 1 to other PWR plants. The functional event trees were developed to describe the system functions that must work or not work in order for a certain accident sequence to happen, one for pipe breaks and one for transients

  13. Study of single particle properties of nuclei in the region of the "island of inversion" by means of neutron-transfer reactions

    CERN Multimedia

    Kruecken, R; Voulot, D

    2007-01-01

    We are aiming at the investigation of single particle properties of neutron-rich nuclei in the region of the "island of inversion" where intruder states from the $\\{fp}$-shell favour deformed ground states instead of the normal spherical $\\textit{sd}$-shell states. As first experiment, we propose to study single particle states in the neutron-rich isotope $^{31}$Mg. The nucleus will be populated by a one-neutron transfer reaction with a $^{30}$Mg beam at 3 MeV/u obtained from REX-ISOLDE impinging on a CD$_{2}$ target. The $\\gamma$-rays will be detected by the MINIBALL array and the particles by a newly built set-up of segmented Si detectors with a angular coverage of nearly 4$\\pi$. Relative spectroscopic factors extracted from the cross sections will enable us to pin down the configurations of the populated states. These will be compared to recent shell model calculations involving new residual interactions. This will shed new light on the evolution of single particle structure leading to the breaking of the ...

  14. Statistical Analysis of the Spatial Distribution of Multi-Elements in an Island Arc Region: Complicating Factors and Transfer by Water Currents

    Directory of Open Access Journals (Sweden)

    Atsuyuki Ohta

    2017-01-01

    Full Text Available The compositions and transfer processes affecting coastal sea sediments from the Seto Inland Sea and the Pacific Ocean are examined through the construction of comprehensive terrestrial and marine geochemical maps for western Japan. Two-way analysis of variance (ANOVA suggests that the elemental concentrations of marine sediments vary with particle size, and that this has a greater effect than the regional provenance of the terrestrial material. Cluster analysis is employed to reveal similarities and differences in the geochemistry of coastal sea and stream sediments. This analysis suggests that the geochemical features of fine sands and silts in the marine environment reflect those of stream sediments in the adjacent terrestrial areas. However, gravels and coarse sands do not show this direct relationship, which is likely a result of mineral segregation by strong tidal currents and the denudation of old basement rocks. Finally, the transport processes for the fine-grained sediments are discussed, using the spatial distribution patterns of outliers for those elements enriched in silt and clay. Silty and clayey sediments are found to be transported and dispersed widely by a periodic current in the inner sea, and are selectively deposited at the boundary of different water masses in the outer sea.

  15. A Credibility-Based Chance-Constrained Transfer Point Location Model for the Relief Logistics Design (Case Study: Earthquake Disaster on Region 1 of Tehran City

    Directory of Open Access Journals (Sweden)

    Ahmad Mohamadi

    2015-02-01

    Full Text Available Occurrence of natural disaster inflicts irreparable injuries and symptoms on humans. In such conditions, affected people are waiting for medical services and relief commodities. Thus, quick reaction of medical services and relief commodities supply play important roles in improving natural disaster management. In this paper, a multi-objective non-linear credibility-based fuzzy mathematical programming model under uncertainty conditions is presented, which considers two vital needs in disaster time including medical services and relief commodities through location of hospitals, transfer points, and location routing of relief depots. The proposed model approaches reality by considering time, cost, failures probability in routes, and parameters uncertainty. The problem is first linearized and then global criterion method is applied for solving the multi objective model. Moreover, to illustrate model efficiency, a case study is performed on region 1 of Tehran city for earthquake disaster. Results demonstrate that if Decision-makers want to meet uncertainty with lowered risk, they have to choose a high minimum constraint feasibility degree even though the objective function will be worse.

  16. Study of PWR reactor efficiency as a function of turbine steam extractions; Estudo da otimizacao da eficiencia de reator PWR em funcao das extracoes de vapor da turbina

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Janine Gandolpho da; Alvim, Antonio Carlos Marques; Martinez, Aquilino Senra [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The objective of this work is to optimize the extractions of the low-pressure turbine of a PWR nuclear reactor, in order to obtain the best thermodynamic cycle efficiency. We have analyzed typical data of a 1300 MW PWR reactor, operating at 25%, 50%, 75% and 100% capacities, respectively. The first stage of this study consists of generating a mathematical model capable of describing the reactor behavior and efficiency at any power level. The second stage of this study consists of to combine the generated mathematical model in an optimization computer program that optimize the extractions flow of the low-pressure turbine until it finds the optimal system efficiency. This work does not alter the nuclear facility project in any way. (author)

  17. 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

  18. PWR Secondary Water Chemistry Control Status: A Summary of Industry Initiatives, Experience and Trends Relative to the EPRI PWR Secondary Water Chemistry Guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, Keith; Choi, Samuel

    2012-09-01

    The latest revision of the EPRI Pressurized Water Reactor (PWR) Secondary Water Chemistry Guidelines was issued in February 2009. The Guidelines continue to focus on minimizing stress corrosion cracking (SCC) of steam generator tubes, as well as minimizing degradation of other major components / subsystems of the secondary system. The Guidelines provide a technically-based framework for a plant-specific and effective PWR secondary water chemistry program. With the issuance of Revision 7 of the Guidelines in 2009, many plants have implemented changes that allow greater flexibility on startup. For example, the previous Guidelines (Revision 6) contained a possible low power hold at 5% power and a possible mid power hold at approximately 30% power based on chemistry constraints. Revision 7 has established a range over which a plant-specific value can be chosen for the possible low power hold (between 5% and 15%) and mid power hold (between 30% and 50%). This has provided plants the ability to establish significant plant evolutions prior to reaching the possible power hold; such as establishing seal steam to the condenser, placing feed pumps in service, or initiating forward flow of heater drains. The application of this flexibility in the industry will be explored. This paper also highlights the major initiatives and industry trends with respect to PWR secondary chemistry; and outlines the recent work to effectively address them. These will be presented in light of recent operating experience, as derived from EPRI's PWR Chemistry Monitoring and Assessment (CMA) program (which contains more than 400 cycles of operating chemistry data). (authors)

  19. PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

    2010-01-01

    An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

  20. Measurement of local flow pattern in boiling R12 simulating PWR conditions with multiple optical probes

    International Nuclear Information System (INIS)

    Garnier, J.

    1998-01-01

    For a comprehensive approach of boiling crisis phenomenon in order to get more reliable predictions of critical heat flux in PWR core, a flow pattern study is under progress at CEA GRENOBLE (in a joint program with Electricite de France: EdF). The first aim is to get experimental results on flow structure in the range of thermal hydraulic parameters involved in the core of a PWR (pressure up to 16 MPa, heat flux about 1 MW/m 2 , mass velocity up to 5000 kg/s/m 2 . As critical heat flux is a local phenomenon and is the result of the flow development, the data has to be measured from the beginning of boiling until boiling crisis, and from the bulk flow until the boundary layer close to the heating walls. Therefore, these results will be useful in modeling not only boiling crisis phenomenon but also condensation in subcooled boiling, coalescence, splitting up, mass and energy transfers at interfaces, and so on. In a first step, the test section is a vertical tube 19.2 mm internal diameter with an axial uniform heat flux over a 3.5m length. The study is performed on the DEBORA loop with Freon 12 as coolant fluid. We assume that basic boiling phenomena (and the knowledge we get about them) only depend on the fluid properties by means of dimensionless parameters but not on the fluid itself. In a first part, we briefly recall that interfacial detection is the most important parameter of a flow pattern study. Therefore, the use of probes able to measure the Phase Indicator Function (P.I.F.) is necessary. A first study of flow conditions shows that the flow pattern is essentially a bubbly one with vapor particles of low diameter (about 300 clm) and high velocity (up to 7 m/s). These criteria induce that a multiple optical probe is the most appropriate tool provided we improve the technology. We detail the way to obtain probes able to detect small particles at high velocity. Each fiber is stretched to get a tip of 10 Clm with the cladding kept on 50 μm length which defines