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Sample records for wh type pwr

  1. The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

    International Nuclear Information System (INIS)

    Ryu, Eun Hyun; Song, Yong Mann

    2014-01-01

    The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H 2 O) and heavy water (D 2 O). Also, it is well known that the slowing-down ratio of D 2 O is hundreds of times larger than that of H 2 O while the slowing-down power of H 2 O is several times larger than that of D 2 O. This means that the H 2 O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each lattice

  2. The Difference between Flux Spectrums of WH-type Assembly and CANDU-type Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Eun Hyun; Song, Yong Mann [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The nuclear reactors are categorized by the material of the moderator because of its importance. The representative materials of the moderator are light water (H{sub 2}O) and heavy water (D{sub 2}O). Also, it is well known that the slowing-down ratio of D{sub 2}O is hundreds of times larger than that of H{sub 2}O while the slowing-down power of H{sub 2}O is several times larger than that of D{sub 2}O. This means that the H{sub 2}O sometimes plays a role of an absorber such as the liquid zone controller (LZC) in a CANDU-type reactor. It is thought that the flux spectrums in a different reactor can differ from each other. In this research, two representative assemblies (the Westinghouse (WH)-type fuel assembly of PWR and the CANDU-type fuel lattice of PHWR) are selected and the flux results for each group are extracted. Although there are many codes for the lattice transport calculation, the WIMS code and the HELIOS code are used for the calculation of the WH-type fuel lattice and the CANDU-type fuel lattice. A clear difference in spectrum between the CANDU-type lattice and WH 16GD-type lattice is confirmed. Because of the superior moderating ratio of the heavy water, the thermal flux ratio of the CANDU-type lattice is almost 82%, while that of the WH 16 GD-type lattice is around 23%. Because of the large portion of the thermal flux in the CANDU-type lattice, the boron effect is maximized with the result from variations of boron. Thus it can be said that the spectrum largely depends on the moderator material, and the boron effect and sensitivity largely depends on the flux spectrum. Because of the dominant effect of the moderator material on the flux spectrum in a nuclear reactor, in the future, a comparison of the spectra of SFR, HTGR, PWR, and PHWR are also an interesting subject to study. Over-moderation in PHWR lattice and under-moderation in PWR lattice can be explained by the investigation about flux spectrums with variations of moderator density in each

  3. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  4. A Comparative analysis for control rod drop accident in WH and CE type

    International Nuclear Information System (INIS)

    Yang, Chang-Keun; Kim, Yo-Han; Ha, Jun-Sang

    2008-01-01

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. And the results were compared with those mentioned in the final safety analysis reports (FSARs) of the plants

  5. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  6. 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

  7. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  8. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  9. 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.)

  10. 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)

  11. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  12. Simulation of the Long period Core Design for WH type of KHNP

    International Nuclear Information System (INIS)

    Jung, Ji-Eun; Moon, Sang-Rae

    2016-01-01

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve

  13. Simulation of the Long period Core Design for WH type of KHNP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ji-Eun; Moon, Sang-Rae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    The current core design of the reactor and the new design of long period based on ANC code are compared here targeting the unit of WH type(Westinghouse nuclear steam supply system) operated by KHNP. The reactor core is composed of 157 fuel assemblies, consisting of a 17×17 array with 264 fuel rods, 24 guide thimbles. To investigate susceptibility of CIPS(crud-induced power shift) for long period core design, the boron mass is also calculated here. The long period core design for WH type of KHNP is simulated and evaluated the risk assessment for the result. 89 feed assemblies and 4.95w/o uranium enrichment (3.2w/o for Axial-blanket) are used for fresh fuel rods. The cycle length of long period design is increased by 6 month than the average of operated cycles satisfying the criteria of risk assessment for the core design; maximum F△h and maximum pin burnup and so on, except burndown curve.

  14. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  15. Hearing dysfunction in heterozygous Mitf(Mi-wh) /+ mice, a model for Waardenburg syndrome type 2 and Tietz syndrome.

    Science.gov (United States)

    Ni, Christina; Zhang, Deming; Beyer, Lisa A; Halsey, Karin E; Fukui, Hideto; Raphael, Yehoash; Dolan, David F; Hornyak, Thomas J

    2013-01-01

    The human deafness-pigmentation syndromes, Waardenburg syndrome (WS) type 2a, and Tietz syndrome are characterized by profound deafness but only partial cutaneous pigmentary abnormalities. Both syndromes are caused by mutations in MITF. To illuminate differences between cutaneous and otic melanocytes in these syndromes, their development and survival in heterozygous Microphthalmia-White (Mitf(Mi-wh) /+) mice were studied and hearing function of these mice characterized. Mitf(Mi-wh) /+ mice have a profound hearing deficit, characterized by elevated auditory brainstem response thresholds, reduced distortion product otoacoustic emissions, absent endocochlear potential, loss of outer hair cells, and stria vascularis abnormalities. Mitf(Mi-wh) /+ embryos have fewer melanoblasts during embryonic development than their wild-type littermates. Although cochlear melanocytes are present at birth, they disappear from the Mitf(Mi-wh) /+ cochlea between P1 and P7. These findings may provide insight into the mechanism of melanocyte and hearing loss in human deafness-pigmentation syndromes such as WS and Tietz syndrome and illustrate differences between otic and follicular melanocytes. © 2012 John Wiley & Sons A/S.

  16. Prompt-Type Frequency, Auditory Pattern Discrimination, and EFL Learners' Production of "WH"-Questions

    Science.gov (United States)

    McDonough, Kim; De Vleeschauwer, Jindarat

    2012-01-01

    Recently researchers have suggested that syntactic priming may facilitate the production of "wh"-questions with obligatory auxiliary verbs, particularly when learners are prompted to produce those questions with a wide variety of lexical items (McDonough & Kim, 2009; McDonough & Mackey, 2008). However, learners' ability to benefit from syntactic…

  17. Wh-question intonation in Peninsular Spanish: Multiple contours and the effect of task type

    Directory of Open Access Journals (Sweden)

    Nicholas C. Henriksen

    2009-06-01

    Full Text Available This paper reports on an experimental investigation of wh-question intonation in Peninsular Spanish. Speech data were collected from six León, Spain Peninsular Spanish speakers, and oral production data were elicited under two conditions: a computerized sentence reading task and an information gap task-oriented dialogue. The latter task was an adaptation of the HCRC Map Task method (cf. Anderson et al., 1991 and was designed to elicit multiple wh-question productions in an unscripted and more spontaneous speech style than the standard sentence reading task. Results indicate that four contours exist in the tonal inventory of the six speakers. The two most frequent contours were a final rise contour and a nuclear circumflex contour. Systematic task-based differences were found for four of the six speakers, indicating that sentence reading task data alone may not accurately reflect spontaneous speech tonal patterns (cf. Cruttenden, 2007; but see also Lickley, Schepman, & Ladd, 2005. The experimental findings serve to clarify a number of assumptions about the syntax-prosody interface underlying wh-question utterance signaling; they also have implications for research methods in intonation and task-based variation in laboratory phonology.

  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. 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)

  20. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  1. 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)

  2. Performance of High Temperature Operational Amplifier, Type LM2904WH, under Extreme Temperatures

    Science.gov (United States)

    Patterson, Richard; Hammoud, Ahmad; Elbuluk, Malik

    2008-01-01

    Operation of electronic parts and circuits under extreme temperatures is anticipated in NASA space exploration missions as well as terrestrial applications. Exposure of electronics to extreme temperatures and wide-range thermal swings greatly affects their performance via induced changes in the semiconductor material properties, packaging and interconnects, or due to incompatibility issues between interfaces that result from thermal expansion/contraction mismatch. Electronics that are designed to withstand operation and perform efficiently in extreme temperatures would mitigate risks for failure due to thermal stresses and, therefore, improve system reliability. In addition, they contribute to reducing system size and weight, simplifying its design, and reducing development cost through the elimination of otherwise required thermal control elements for proper ambient operation. A large DC voltage gain (100 dB) operational amplifier with a maximum junction temperature of 150 C was recently introduced by STMicroelectronics [1]. This LM2904WH chip comes in a plastic package and is designed specifically for automotive and industrial control systems. It operates from a single power supply over a wide range of voltages, and it consists of two independent, high gain, internally frequency compensated operational amplifiers. Table I shows some of the device manufacturer s specifications.

  3. Method of stopping operation of PWR type reactor

    International Nuclear Information System (INIS)

    Ueno, Takashi; Tsuge, Ayao; Kawanishi, Yasuhira; Onimura, Kichiro; Kadokami, Akira.

    1989-01-01

    In PWR type reactors after long period of l00 % power operation, since boiling is caused in heat conduction pipes and water is depleted within the intergranular corrosion fracture face in the crevis portion to result in a dry-out state, impregnation and concentration of corrosion inhibitors into the intergranular corrosion fracture face are insufficient. In view of the above, the corrosion inhibitor at a high concentration is impregnated into the intergranular corrosion fracture face by keeping to inject the corrosion inhibitor from l00 % thermal power load by way of the thermal power reduction to the zero power state upon operatioin shutdown. That is, if the thermal power is reduced to or near the 0 power upon reactor shutdown, feedwater in the crevis portion is put to subcooled state, by which the steam present in the intergranular corrosion fracture face are condensated and the corrosion inhibitor at high concentration impregnated into the crevis portion are penetrated into the intergranular corrosion fracture face. (K.M.)

  4. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  5. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  6. 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

  7. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  8. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  9. Programme of hot points eradication (Co-60) led on French PWR type reactors

    International Nuclear Information System (INIS)

    Rocher, A.; Ridoux, P.; Anthoni, S.; Brun, C.

    1998-01-01

    The question of hot points (pellets rich in cobalt 59 or in cobalt 60 in a PWR type reactor), is studied from the radiation protection point of view. The purpose is to see how to optimize the radiation protection, the elimination of these hot points can bring an improvement. (N.C.)

  10. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  11. Comparative study T-type and I-type layout of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Eko Rudi Iswanto and Siti Alimah

    2010-01-01

    Determining plant layout is one of the five major stages during the life time of a nuclear power plant. Some important factors that affect in the selecting of plant layout are availability of infrastructure, economic aspects, social aspects, public and environment safety, and also easy to do. Another factor to be considered is requirements as seismic design, which refers to the principles of good security workers, communities and the environment of radiological risks. There are many layout types of nuclear power plant, two of them are T-type layout and I-type layout. Each type of the plant layout has advantage and disadvantage, therefore this study is to understand them. Good layout is able to provide a high level of security against earthquakes. In term of earthquake design, I-type layout has a higher security level than T-type layout. Therefore, I-type layout can be a good choice for PWR nuclear power plants 1000 MWe that will be built in Indonesia. (author)

  12. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    Drab, F.; Grof, V.

    1978-01-01

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  13. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  14. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  15. Study of crack propagation velocity in steel tanks of PWR type reactor

    International Nuclear Information System (INIS)

    Amzallac, C.; Bernard, J.L.; Slama, G.

    1983-05-01

    Description and results of a serie of tests carried out on crack propagation velocity of steels in PWR environment (pressurized high temperature water), in order to examine the effects of metallurgical parameters such as chemical composition of steel, especially sulfur and carbon content, and steel type (laminate or forged steels), effects of mechanical parameters such as loading ratio, cycle form, frequency and application mode of loads and of chemical parameters (anodal dissolution or fatigue with hydrogen) [fr

  16. Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

    International Nuclear Information System (INIS)

    Barcellos, C.S. de.

    1980-01-01

    Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

  17. Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

    1979-01-01

    The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

  18. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  19. New generation nuclear power units of PWR type integral reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

    1997-01-01

    Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

  20. Technical report: fabrication of PWR type rodlet fuel

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Uno, Hisao; Sasajima, Hideo

    1990-06-01

    With respect to the simulated reactivity initiated accident (RIA) experiments with pre-irradiated LWR type fuel rods at nuclear safety research reactor (NSRR), there were principally three technical difficulties which should be overcome: (1) Fabrication of the rodlet fuel; Fuel rods from the commercial power reactors had an active column length by 3.6m. To utilize this for NSRR pulse experiment, rodlet fuel having an active column length by 0.12m (reduced to one thirtieth) is requested to fabricate without changing the inside fuel conditions. (2) Development of in-core instrumentations: During pre-irradiation stages, a long-sized fuel rod had dimensional changes by waterside corrosion, bowing, creep down and so on. The fuel also had greater amount of radioactive fission products. This condition is significant to in-core instrumentations to be attached to the fuel rods. Well characterized data to be obtained from these, however, are quite necessary and important from research point of view. Remote handling techniques to attach the rod pressure sensor, the cladding extensometer, the fuel extensometer, and the cladding surface thermocouple to pre-irradiated fuel rods are, therefore, requested to develop. (3) Installation of PIE equipments for pulsed rodlet fuels: PIE on the pulsed rodlet fuels are necessary to better understanding the fuel performance detaily. Equipments which can easily detect the data related to PCMI type fuel failure are matter of concern. Since 1986, the technical difficulties have been tried to overcome by all staffs belonging to Reactivity Accident Laboratory, NSRR Operation Division, Department of Reactor Fuel Examination and Hot Laboratory. This report describes the technical achievements obtained through four years work. (author)

  1. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

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

  3. An intelligent pedagogic tool for teaching the operators of PWR type reactors

    International Nuclear Information System (INIS)

    Cordier, B.; Guillermard, M.

    1990-01-01

    A tool was developed for assisting the instruction of the operators of a PWR type nuclear power plant. For achieving the objectives, an expert system and a simulator were combined. The main objective of the system is to improve the work of the operators in performing remedial actions in case of accident. The simulator applies two IBM PC AT3 and a MC 680 20 microprocessor. The use and the validation of the expert system are presented. The perspectives for the system, implanted on the Tricastin nuclear power plant, are analyzed [fr

  4. Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

    International Nuclear Information System (INIS)

    Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

    2015-01-01

    Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

  5. Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    2000-03-01

    At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

  6. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR

    International Nuclear Information System (INIS)

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G.

    2008-01-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO 2 . (Author)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  8. Design of the control room of the N4-type PWR: main features and feedback operating experience

    International Nuclear Information System (INIS)

    Peyrouton, J.M.; Guillas, J.; Nougaret, Ch.

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1994-12-31

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

  11. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  12. PWR type overpower tests at 1620 GJ/KGU (18,800 MWD/MTU)

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.; Carlsen, H.

    1979-01-01

    Three PWR type test fuel pins accumulated a burnup of 1620 GJ/kgU at heat loads decreasing from 45 to 28 kW/m (avg. test levels). One pin was ramped to 43 kW/m at 31 W/m/s; after 15 ks the power was increased to 45 kW/m and held constant for 1.9 Ms without failure indication. The other two pins were ramped to 44 kW/m at 23 W/m/s and then to 49 kW/m in a further 1.2 ks; both failed after max. 360 s. The post-irradiation examination revealed large stress-corrosion (SCC) type cladding cracks. Other cracks, down to a few μm deep, were probably early stages of large SCC cracks. Fission gas release in the intact pin was as high as 42% and estimated to be much lower for the two failed pins

  13. Probing the WL'WH and WR'WH interactions at LHC

    International Nuclear Information System (INIS)

    Bao Shoushan; Li Honglei; Si Zongguo; Zhou Yufeng

    2011-01-01

    Many new physics models predict the existence of TeV-scale charged gauge boson W ' together with Higgs-boson(s). We study the W ' WH interaction and explore the angular distribution of charged leptons to distinguish W R ' WH from W L ' WH in the pp→HW→bblν process at the LHC. It is found that a new type forward-backward asymmetry (A FB ) relating to the angle between the direction of the charged lepton in the W rest frame and that of the reconstructed W ' in the laboratory frame is useful to investigate the properties of W ' WH interaction. We analyze the standard model backgrounds and develop a set of cuts to highlight the signal and suppress the backgrounds at LHC. We find that A FB can reach 0.03(-0.07) for W R ' (W L ' ) production at √(S)=14 TeV.

  14. Wh-optatives in Danish

    DEFF Research Database (Denmark)

    Ørsnes, Bjarne

    2013-01-01

    Danish has an optative taking the form of a wh-clause with embedded word order: Hvem der var rig! lit. ‘Who there were rich!’ with the intended meaning Wish I were rich! Such wh-clauses cannot be used as optatives in German or English (or only marginally so). On the basis of a detailed analysis...... of the syntax, semantics, and pragmatics of this little-studied phenomenon, this article seeks to explain why a wh-clause with “embedded” word order can be used as an optative. I argue that wh-optatives are compositional in many but not all respects. It is a special construction whose interpretation...... is determined to a large extent by its grammatical properties but which also has completely idiosyncratic properties. In German and English, such wh-clauses are only potential optatives: They are not used as optatives, but nothing would preclude it....

  15. Vibration characteristics of a PWR fuel rod supported by optimized H type spacer grids

    International Nuclear Information System (INIS)

    Choi, M. H.; Kang, H. S.; Yoon, K. H.; Kim, H. K.; Song, K. N.

    2002-01-01

    The spacer grids are one of the main structural components in the fuel assembly, which supports and protects the fuel rods from the external loads by seismic and coolant flow. In this study, a modal test and a FE vibration analysis using ABAQUS are performed on a PWR dummy fuel rod of 3.847 m which is continuously supported by eight Optimized H type spacer grids. The experimental results agree with previous works that the natural frequencies decrease, while the amplitudes increase, with the increase of the excitation force. The force levels showing the maximum displacement of 0.2 mm are in the range from 0.2 N to 0.3 N, and at the same force range the fundamental frequencies are measured around 42.0 Hz, at which the relatively big displacements are observed at the 7th span. The results from the modal tests and the FE analyses are compared by both Modal Assurance Criteria (MAC) values and mode shapes. The MAC values at 2nd, 4th, and 7th mode are below 50%. It is believed that the reason of the low MACs at those modes is that the vibration amplitudes of the modes are more distorted by the excitation force than those of the other modes

  16. Reliability analysis of 2 types of auxiliary feedwater system for PWR

    International Nuclear Information System (INIS)

    Ekariansyah, Andi Sofrany

    2002-01-01

    This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

  17. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  18. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Belem, J.A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  19. Evaluation 2 of B10 depletion in the WH PWR

    International Nuclear Information System (INIS)

    Park, Sang Won; Woo, Hae Suk; Kim, Sun Doo; Chae, Hee Dong; Myung, Sun Yup; Jang, Ju Kyung

    2001-01-01

    This paper presents the methodology to evaluate the B 10 depletion behavior in the pressurized water reactor. And B 10 depletion evaluation is performed based on the prediction program and the measured data of B 10 . The result shows that B 10 depletion during normal operation is not negligible. Therefore, adjustments for this depletion effect should be made to calculate the estimated critical postion(ECP) and determine the boron concentration required to maintain the specified shutdown margin

  20. 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

  1. Results of safety analysis on PWR type nuclear power plants with two and three loops

    International Nuclear Information System (INIS)

    1979-01-01

    The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

  2. Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

    International Nuclear Information System (INIS)

    Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

    2008-01-01

    As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

  3. Operation results of the secondary circuits of the French PWR type power plant park

    International Nuclear Information System (INIS)

    Mercier, J.P.

    1984-01-01

    Global results of performances realized since 1981 by the French PWR 900 MW power plants (installed power, availability, casual or planned shutdowns); analysis of the behaviour (casual unavailability) comparing together the performances of the different components in the secondary circuit; behaviour of the principal materials of the secondary circuit and their weight in the unavailabilities of the whole French nuclear park [fr

  4. Effect of operating conditions and environment on properties of materials of PWR type nuclear power plant components

    International Nuclear Information System (INIS)

    Vacek, M.

    1987-01-01

    Operating reliability and service life of PWR type nuclear power plants are discussed with respect to the material properties of the plant components. The effects of the operating environment on the material properties and the methods of their determination are characterized. Discussed are core materials, such as fuel, its cladding and regulating rod materials, and the materials of pipes, steam generators and condensers. The advances in the production of pressure vessel materials and their degradation during operation are treated in great detail. (Z.M.)

  5. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

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

    International Nuclear Information System (INIS)

    Mansouri, S.

    1985-06-01

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

  7. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  8. Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

    International Nuclear Information System (INIS)

    Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

    1997-01-01

    Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

  9. Pre design processing of waste of ex-resin without materials matrix from nuclear power plant type PWR 1000 MW

    International Nuclear Information System (INIS)

    Cerdas Tarigan

    2010-01-01

    Have been done pre design processing of waste ex-resin without capacities matrix materials from nuclear power plant type PWR 1000 MW During the time radioactive waste of ex-resin processed to use process of immobilization use matrix materials like mixture cement and epoxy resin and then conditioning. This process is not effective and efficient because end result volume of end product bigger than volume early operation system and maintenance of its installation more difficult. To overcome this created a design of technology processing of waste of ex- resin without matrix materials through process of strainer, drying and conditioning represent technological innovation newly processing of radioactive waste of ex-resin. Besides this process more effective and efficient, volume of end product waste much more small from volume early and operation system and maintenance of its easier installation. Pre design is expected to be used as a basis to make conceptual of pre design installation of strainer, drying and conditioning for the processing of waste of ex-resin from nuclear power plant type PWR 1000 MW. (author)

  10. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  11. Re-irradiation and limit testing of the fuels PWR type reactors

    International Nuclear Information System (INIS)

    Roche, M.; Molvault, M.

    1978-01-01

    In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

  12. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

  13. 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

  14. A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

    International Nuclear Information System (INIS)

    Rossini, M.R.

    1992-01-01

    An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

  15. Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-01-01

    Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

  16. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Belem, J A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

  17. 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

  18. Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

    International Nuclear Information System (INIS)

    Castano, M. L.; Blazquez, F.; Gomez Briceno, D.; Lagares, A.

    1998-01-01

    The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

  19. 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

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is 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. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  20. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  1. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  2. Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

    1985-01-01

    The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

  3. Preventive testing and leakage detection in pipe-lines of steam condensers and generators of a PWR type reactor

    International Nuclear Information System (INIS)

    Canalini, A.; Carvalho, N.C. de

    1985-01-01

    The non-destructive methods: Spum, Helium and Hydrostatic used in leakage detection in condenser pipelines for PWR type reactors are presented. The time, costs, sensitivity, resources necessary and personnel development factors are considered to choose adequated method, in function of nuclear power plant conditions. The leakage tests are applied in pressurized systems or vacuum. Eddy Current testing is used in condensers and steam generators aiming to avoid leakage in these equipments. The spume testing for leakage detection in condenser pipelines - which operation - and hydrostatic testing for leakage detection through reaming with shutdown - were most efficients. The Helium testing applied in pressurized systems or submitted to vacuum systems presented satisfactory results. The Eddy Current testing in condenser and steam generator pipelines reached desired objective, reducing leakage in the first and preserving the integrity in the second. (M.C.K.) [pt

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Science.gov (United States)

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

    2018-02-01

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

  7. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

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

  8. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors

    International Nuclear Information System (INIS)

    Lantes, B.; Bienvenu, Ph.

    2001-01-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  9. Evaluation of passive autocatalytic recombiners (PARS) performance for a PWR-konvoi containment type with Gothic 8.1 code

    International Nuclear Information System (INIS)

    Lopez-Alonso Conty, E.; Papini, D.; Jimenez Varas, G.

    2016-01-01

    The study presented in this work analyses the evaluation of Passive Autocatalytic Recombiners (PARs) performance for a PWR-Konvoi containment type as a result of an international collaboration between the Paul Scherrer institute (PSI) and the Universidad Politecnica de Madrid (UPM). The implementation study analyzes the size, location and number of the PARs to minimize the risk arising from a hydrogen release and its distribution in the containment building during a hypothetical severe accident. A detailed 3D model of containment was used for the simulations developed for the Gothic 8.1 code. In the first place, the hydrogen preferential pathways and points of hydrogen accumulation were studies and identified starting from the base case scenario without any mitigation measure. The severe accident scenario chosen is a fast release of hydrogen-steam mixture from hot leg creep rupture during SBO (Station Black-Out) accident. Secondly a configuration of PARs was simulated under the same conditions of the unmitigated case. The PAR configuration offered an improvement in the chosen accident scenario, decreasing the hydrogen concentration values below the flammability limit /hydrogen concentration below 7%) in all the containment compartments. (Author)

  10. "Wh"-Questions in the English Language

    Science.gov (United States)

    Mowarin, Macaulay

    2011-01-01

    This paper analyzes "wh"-questions in the English Language based mainly on Chomsky's Minimalist Programme of transformational grammar as the theoretical model. The four main objectives of this paper are as follows: first, it undertakes a cross linguistic typological analysis of "wh"-questions and it then discusses the derivation of…

  11. Development of fabrication process of upper nozzle BIBLIS type of PWR fuel element

    International Nuclear Information System (INIS)

    Miranda, O.; Lorenzo, D.F.R.

    1982-01-01

    Process and parameters of milling and welding of a upper nozzle BIBLIS type prototype are presented. Milling process, cutting tools studies, production devices and inspection were developed and researched. (author) [pt

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

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

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

  13. Development of an engineering simulator for integral type PWR for nuclear ship

    International Nuclear Information System (INIS)

    Takahashi, Teruo; Shimazaki, Junya; Nakazawa, Toshio

    2000-01-01

    JAERI has developed a real-time engineering simulator for the integral type reactor MRX (Marine Reactor X) of power 100 MWt to evaluate the design and operational performance and to study highly automatic operations of a reactor plant. Marine reactor is operated under the conditions of pitching and rolling and load change, in comparison with a reactor for a land-based generating plant. And the MRX has systems with structural features, such as water-filled containment vessel, once-through type steam generator and emergency decay heat removal system. Considerations are paid to take these operational conditions and structural features into the simulation model. It is shown that the simulated results are consistent with the planned design and operational performance, and on the other hand present us some technical issues to be investigated in the design specifications. (author)

  14. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1985-02-01

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

  15. Study of behaviour of radioactive iodine inorganic compounds in PWR type reactor loops

    International Nuclear Information System (INIS)

    Alm, M.; Johannsen, K.-H.; Dreyer, R.

    1980-01-01

    Compounds of radioactive iodine and its distribution between water and vapour depending on temperature, pressure and water regime of reactor coolant with water under pressure are investigated. The field of variation of parameters indicated is widened as compared with operating reactor parameters (pressure 2-14 MPa, temperature 210-335 deg C). Distribution of iodine compounds has been studied by a statistical method. For WWER-type reactors the following conclusions have been drawn: radioactive iodine in water and vapor in the first and second loops exists in the form of iodide, radioactive iodine concentration in water vapour at constant temperature and pressure mainly is depended on water pH value, radioactive iodine solubility in water vapor at normal parameters of the reactor first loop can be approximately calculated by the equation: Ksub(d)=Csub(g)/Csub(l)=(rhosub(g)/rhosub(l))sup(2), where Ksub(d) is a coefficient of solid distribution between water and vapour, rho is density c is concentration [ru

  16. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

  17. 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)

  18. 48 CFR 53.303-WH-347 - Department of Labor Form WH-347, Payroll (For Contractor's Optional Use).

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Department of Labor Form WH-347, Payroll (For Contractor's Optional Use). 53.303-WH-347 Section 53.303-WH-347 Federal... Illustrations of Forms 53.303-WH-347 Department of Labor Form WH-347, Payroll (For Contractor's Optional Use...

  19. Towards a PSA harmonization French-Belgian comparison of the level 1 PSA for two similar PWR types

    International Nuclear Information System (INIS)

    Dupuy, P.; Corenwinder, F.; Lanore, J.M.; Gryffroy, D.; Gelder, P. de; Hulsmans, M.

    2002-06-01

    In the framework of the cooperation between French and Belgian regulatory authorities, a PSA (Probabilistic Safety Assessment) comparison exercise has been carried out for several years. This comparison deals with two PSA level 1 studies for internal events, performed for both power and shutdown states: the French PSA of the 900 MWe-series PWR, and the Belgian PSA of the Tihange 1 PWR, which both concern PWRs with a similar Framatome design. The purpose of this paper is to describe the PSA comparison methodology and to present, in a qualitative way, an overview of the insights obtained up to now. It also shows that such an 'a posteriori' benchmark exercise turns out to be a step towards PSA harmonization, and gives more confidence in the results of plant specific PSA when used for applications like precursor analysis or evaluations of importance to safety. (authors)

  20. Expert system for assisting the repair operations on the control racks of the control rods assembly in a 900 MW PWR type reactor

    International Nuclear Information System (INIS)

    Monnier, B.; Doutre, J.L.; Franco, A.

    1990-01-01

    The expert system presented was developed for assisting the repair operations on the control equipment of the control rod assembly in a PWR type reactor. The expert system allows the representation of expert knowledge and diagnostic reasoning. The objective of the expert system is to achieve the most precise diagnostic and localizing of the breakdown elements, by processing the data acquired during breakdown. The development steps, the structure and the applications of the expert system are summarized. The expert system operates in an IBM PC equipped with a AMAIA 8 Mo card. A time schedule of 18 months is predicted [fr

  1. Application of the perturbation theory-differential formalism-for sensitivity analysis in steam generators of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Sanders, R.M.G.; Andrade Lima, F.R. de; Alvim, A.C.M.

    1987-06-01

    An homogeneous model which simulates the stationary behavior of steam generators of PWR type reactors and uses the differential formalism of perturbation theory for analysing sensibility of linear and non-linear responses, is presented. The PERGEVAP computer code to calculate the temperature distribution in the steam generator and associated importance function, is developed. The code also evaluates effects of the thermohydraulic parameter variation on selected functionals. The obtained results are compared with results obtained by GEVAP computer code . (M.C.K.) [pt

  2. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  3. Do Persian Native Speakers Prosodically Mark Wh-in-situ Questions?

    Science.gov (United States)

    Shiamizadeh, Zohreh; Caspers, Johanneke; Schiller, Niels O

    2018-02-01

    It has been shown that prosody contributes to the contrast between declarativity and interrogativity, notably in interrogative utterances lacking lexico-syntactic features of interrogativity. Accordingly, it may be proposed that prosody plays a role in marking wh-in-situ questions in which the interrogativity feature (the wh-phrase) does not move to sentence-initial position, as, for example, in Persian. This paper examines whether prosody distinguishes Persian wh-in-situ questions from declaratives in the absence of the interrogativity feature in the sentence-initial position. To answer this question, a production experiment was designed in which wh-questions and declaratives were elicited from Persian native speakers. On the basis of the results of previous studies, we hypothesize that prosodic features mark wh-in-situ questions as opposed to declaratives at both the local (pre- and post-wh part) and global level (complete sentence). The results of the current study confirm our hypothesis that prosodic correlates mark the pre-wh part as well as the complete sentence in wh-in-situ questions. The results support theoretical concepts such as the frequency code, the universal dichotomous association between relaxation and declarativity on the one hand and tension and interrogativity on the other, the relation between prosody and pragmatics, and the relation between prosody and encoding and decoding of sentence type.

  4. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  5. Operating function tests of the PWR type RHR pump for engineering safety system under simulated strong ground excitation

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, Kazuhiro; Homma, Toshiaki; Inazuka, Hisashi; Nakajima, Norifumi.

    1979-08-01

    Results are described of operating function verification tests of a PWR RHR pump during an earthquake. Of the active reactor components, the PWR residual heat removal pump was chosen from view points of aseismic classification, safety function, structural complexity and past aseismic tests. Through survey of the service conditions and structure of this pump, seismic test conditions such as acceleration level, simulated seismic wave form and earthquake duration were decided for seismicity of the operating pump. Then, plans were prepared to evaluate vibration chracteristics of the pump and to estimate its aseismic design margins. Subsequently, test facility and instrumentation system were designed and constructed. Experimental results could thus be acquired on vibration characteristics of the pump and its dynamic behavior during different kinds and levels of simulated earthquake. In conclusion: (1) Stiffeners attached to the auxiliary system piping do improve aseismic performance of the pump. (2) The rotor-shaft-bearing system is secure unless it is subjected to transient disturbunces having high frequency content. (3) The motor and pump casing having resonance frequencies much higher than frequency content of the seismic wave show only small amplifications. (4) The RHR pump possesses an aseismic design margin more than 2.6 times the expected ultimate earthquake on design basis. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  7. 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.)

  8. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method

    International Nuclear Information System (INIS)

    Silva, Alonso F.O.

    2007-01-01

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

  9. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Huber, Horacio

    1989-01-01

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

  10. Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2014-01-01

    The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

  11. Water-hammer experimental set-up and water-hammer experimental study for new types of check valve applied to PWR

    International Nuclear Information System (INIS)

    Liu Hanxun.

    1985-01-01

    This paper describes a self-designed constant temperature water-hammer shock test rig with stainless steel loop in which deionized water is used as working medium. To conduct water-hammer shock simulation tests for the countercurrent phenomenon occurred in the process of shutting, stopping, parallelling and switching the coolant loops of nuclear reactor, a specially designed four-way switching valve and its pneumatic mechanism are used. Water-hammer experimental study is performed for two types of PWR's nonshock check valve with diameter of 150 mm and 200 mm simultaneously. Transient performance of the shock waves, magnitude of their peaks and durations of their fluctuation, is obtained. Some analyses for existing calculational method on water-hammer are made

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  13. Economic targets for small PWR reactor designs

    International Nuclear Information System (INIS)

    Board, J.

    1991-01-01

    Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

  14. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de.

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  15. Validation of the REL2005 code package on Gd-poisoned PWR type assemblies through the CAMELEON experimental program

    International Nuclear Information System (INIS)

    Blaise, Patrick; Vidal, Jean-Francois; Santamarina, Alain

    2009-01-01

    This paper details the validation of Gd-poisoned 17x17 PWR lattices, through several configurations of the CAMELEON experimental program, by using the newly qualified REL2005 French code package. After a general presentation of the CAMELEON program that took place in the EOLE critical Facility in Cadarache, one describes the new REL2005 code package relying on the deterministic transport code APOLLO2.8 based on characteristics method (MOC), and its new CEA2005 library based on the latest JEFF-3.1.1 nuclear data evaluation. For critical masses, the average Calculation-to-Experiment C/E's on the k eff are (136 ± 80) pcm and (300 ± 76) pcm for the reference 281 groups MOC and optimized 26 groups MOC schemes respectively. These values include also a drastic improvement of about 250 pcm due to the change in the library from JEF2.2 to JEFF3.1. For pin-by-pin radial power distributions, reference and REL2005 results are very close, with maximum discrepancies of the order of 2%, i.e., in the experimental uncertainty limits. The Optimized REL2005 code package allows to predict the reactivity worth of the Gd-clusters (averaged on 9 experimental configurations) to be C/E Δρ(Gd clusters) = +1.3% ± 2.3%. (author)

  16. From fundamental mode to the PWR type reactors blow off: physical analysis and contribution to the qualification of calculation tools

    International Nuclear Information System (INIS)

    Maghnouj, A.

    1996-01-01

    The work reported in this thesis centres on the resolution of reactor physics problems posed by the use in pressurised water reactors of fuel assemblies containing mixed uranium-plutonium oxide fuel (MOX). The work is essentially dependent on the results of the EPICURE experimental programme carried out between 1988 and 1994 in the reactor EOLE at the Cadarache Research Centre of the CEA. Our contribution to the validation of the computer program APOLLO2 and of its nuclear data library CEA93 shows that this code system satisfactorily calculates the neutronic characteristics of PWR cores. The validation of the experiments has provided useful information concerning the modifications required to be made to the library CEA93, which is based on the basic library of evaluated nuclear data, JEF2. This approach should now be extended to a wider basis of reactor experimental data. The studies of methods for calculating coolant voiding coefficients has made it possible to select suitable methods based on the available deterministic methods of transport theory in 2 ad 3 dimensions. These schemes have given results in satisfactory agreement with the measurements made in EPICURE programme for both local and total coolant voiding. It would now be worth while to validate the chosen methods by comparisons with calculations made using continuous energy Monte Carlo methods. (author)

  17. Whānau Ora; He Whakaaro Ā Whānau: Māori Family Views of Family Wellbeing

    Directory of Open Access Journals (Sweden)

    Amohia F. Boulton

    2014-04-01

    Full Text Available This article presents the findings from two studies that investigated the concept of whānau ora (family wellbeing: One examined the nature of resilience for Māori whānau and how resilience relates to whānau ora; while the second investigated the impact of the Working for Families policy on Māori families’ perceptions of whānau ora. In each study, Māori were asked to define whānau ora for their family. The responses to the “whānau ora” definition question in each of the studies were separated out to derive a unique dataset of 46 whānau definitions of whānau ora. A secondary analysis of responses was undertaken specifically for this article and these were compared to the whānau ora outcome definition outlined in the Report of the Taskforce on Whānau-Centred Initiatives (Taskforce on Whānau-Centred Initiatives, 2010. The degree of concordance between the definitions of whānau ora expressed by Māori families and those espoused by the government’s Taskforce is outlined. The article discusses the variability in understandings around whānau ora and the implications of our analysis for social service delivery and social policy development.

  18. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  19. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  20. Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

  1. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  2. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  3. Wh-filler-gap dependency formation guides reflexive antecedent search

    Directory of Open Access Journals (Sweden)

    Michael eFrazier

    2015-10-01

    Full Text Available Prior studies on online sentence processing have shown that the parser can resolve non-local dependencies rapidly and accurately. This study investigates the interaction between the processing of two such non-local dependencies: wh-filler-gap dependencies (WhFGD and reflexive-antecedent dependencies. We show that reflexive-antecedent dependency resolution is sensitive to the presence of a WhFGD, and argue that the filler-gap dependency established by WhFGD resolution is selected online as the antecedent of a reflexive dependency. We investigate the processing of constructions like (1, where two NPs might be possible antecedents for the reflexive, namely which cowgirl and Mary. Even though Mary is linearly closer to the reflexive, the only grammatically licit antecedent for the reflexive is the more distant wh-NP, which cowgirl. 1. Which cowgirl did Mary expect to have injured herself due to negligence?Four eye-tracking text-reading experiments were conducted on examples like (1, differing in whether the embedded clause was non-finite (1 and 3 or finite (2 and 4, and in whether the tail of the wh-dependency intervened between the reflexive and its closest overt antecedent (1 and 2 or the wh-dependency was associated with a position earlier in the sentence (3 and 4.The results of Experiments 1 and 2 indicate the parser accesses the result of WhFGD formation during reflexive antecedent search. The resolution of a wh-dependency alters the representation that reflexive antecedent search operates over, allowing the grammatical but linearly distant antecedent to be accessed rapidly. In the absence of a long-distance WhFGD (Exp. 3 and 4, wh-NPs were not found to impact reading times of the reflexive, indicating that the parser's ability to select distant wh-NPs as reflexive antecedents crucially involves syntactic structure.

  4. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    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.)

  5. The acquisition of wh-questions in Russian

    Directory of Open Access Journals (Sweden)

    Nina Rojina

    2005-02-01

    Full Text Available In this paper, I study the acquisition of wh-questions in Russian by one Russian-speaking child from an early stage of syntactic development. For my analysis, I follow Plunkett (1992:73 in adopting a weak version of the Continuity Hypothesis where all the principles of UG are available from the outset of the acquisition process, but the structure of the functional template must be built up before parameter setting can begin. I adopt this hypothesis due to its compatibility with my finding that in the acquisition of Russian wh-questions, the functional projection CP appears at a later stage of syntactic development than other functional projections. Following Plunkett, I assume that in the beginning, the landing site for the wh-word is [Spec,TP] and later, for the valuation of other features (such as finiteness, the child is forced to acquire another functional projection (CP in order to create a landing site for the wh-word. The analysis shows that the first wh-questions acquired by the child are wh-complement questions. The child starts to produce subject wh-questions when the CP projection is acquired, and by the age of 2;4.14, the child’s structure of wh-questions is close to that of adult speech.

  6. Agrammatic Comprehension of Wh-Extraction in Danish, English and French

    DEFF Research Database (Denmark)

    Nyvad, Anne Mette

    2010-01-01

    This dissertation discusses the comprehension deficit in agrammatic aphasia and how it relates to wh-extraction. Ostensibly, Broca’s aphasics are unable to understand certain types of syntactic movement structure when the semantic and pragmatic cues are withheld. The present study seeks to examine...... whether results hitherto obtained in crosslinguistic research on wh-extraction, specifically in English and French, also pertain to Danish agrammatic comprehenders, but it also widens the scope of interest: Biclausal structures involving wh-questions have not yet been subject to a systematic investigation......) widely-held account of the comprehension patterns in agrammatism. Focus is on comprehension, but data from research on agrammatic production will be offered when relevant to the theoretical discussion....

  7. Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR. Verification of the simulation results

    Energy Technology Data Exchange (ETDEWEB)

    Farahani, Aref Zarnooshe [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club

    2017-07-15

    Development of a steady-state model is the first step in nuclear safety analysis. The developed model should be qualitatively analyzed first, then a sensitivity analysis is required on the number of nodes for models of different systems to ensure the reliability of the obtained results. This contribution aims to show through sensitivity analysis, the independence of modeling results to the number of nodes in a qualified MELCOR model for a Westinghouse type pressurized power plant. For this purpose, and to minimize user error, the nuclear analysis software, SNAP, is employed. Different sensitivity cases were developed by modification of the existing model and refinement of the nodes for the simulated systems including steam generators, reactor coolant system and also reactor core and its connecting flow paths. By comparing the obtained results to those of the original model no significant difference is observed which is indicative of the model independence to the finer nodes.

  8. 29 CFR Appendix B to Part 825 - Certification of Health Care Provider (Forms WH-380E & WH-380F)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 3 2010-07-01 2010-07-01 false Certification of Health Care Provider (Forms WH-380E & WH-380F) B Appendix B to Part 825 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR OTHER LAWS THE FAMILY AND MEDICAL LEAVE ACT OF 1993 Pt. 825, App. B Appendix B to Part...

  9. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model; Aplicacao da teoria de perturbacao para calculos de sensibilidade em nucleos de reatores PWR, usando um modelo de dois canais

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, A.C.J.G. de

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs.

  10. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  11. Enlarged level-1 PSA in regard to assessment of cross-cutting effects of hazards and consideration of their uncertainties for a KONVOI type PWR reference plant

    International Nuclear Information System (INIS)

    Haider, C.; Hofer, E.; Kloos, M.; Kuntze, W.; Liemersdorf, H.; Roewekamp, M.; Schwinges, B.; Tuerschmann, M.; Brenig, H.W.; Sommerfeld, H.

    2001-01-01

    In the frame of supporting BMU regarding generic questions concerning probabilistic safety analyses for nuclear power plants as well as regarding evaluation of nuclear specific standards and guidelines the significant contributions to damage states resulting from plant internal and external hazards had to be estimated for a German Konvoi type PWR reference plant. Furthermore, the suitability of the available methods for assessing these hazards should be checked. In the report presented hereafter, only the plant internal hazard Fire out of all the hazards to be considered was probabilistically analysed in detail First of all, screening analyses were carried out for identifying relevant plant areas and to assess their respective efficiency for a proper selection procedure. For a selected, plant area identified to be relevant (area of the cable distributions inside the reactor containment) an indepth analysis was performed. This analysis included all the steps of the probabilistic assessment, starting from the estimation of the fire occurrence frequency, followed by investigations on the fire effects and fire propagation, up to the determination of the fire induced failure probabilities of safety related equipment including the consequences on systems. In addition, the analyses contained particular uncertainty and sensitivity studies, for which aleatoric and epistemic uncertainties were distinguished. As a result of the screening analyses as well as of the in-depth investigations regarding the fire hazard, no significant contributions for the total frequencies for system, core, or plant damage states have been found. In this context, it has to be noticed that the study presented hereafter does not cover a complete fire PSA. With respect to assessing the available methods it has been found that improvements concerning the screening process as well as concerning the probabilistic fire event tree analyses are necessary. With regard to further hazards, a site specific

  12. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Energy Technology Data Exchange (ETDEWEB)

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  13. Experience feedback examination in PWR type reactors operating for the 1997-1999 period; Examen du retour d'experience en exploitation des reacteurs a eau sous pression pour la periode 1997-1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The present report is relative to the examination that the permanent group has made on the experience feedback in operation for PWR type reactors for the period 1997-1999 that was on eleven themes chosen by the Nuclear Safety and Radiation Protection Authority. It used analysis reports made by I.R.S.N. in support of four meetings of the permanent group devoted to this examination from April 2001 to June 2002. The different themes were operating uncertainties, machining to vibrations, analysis of incidents and gaseous releases, circuits, human factors, behaviour of electric batteries, risk of cold source loss. (N.C.)

  14. Announcement of recommendations of the Reaktor-Sicherheitskommission. As of 24 July 1997. Joint recommendations of RSK and GPR for safety requirements of future nuclear PWR-type power plants. English versions published in the years 1995 through 1997

    International Nuclear Information System (INIS)

    1997-01-01

    The recommendations, most parts given in English, refer to the European Pressurized Water Reactor (EPR) and have been established by the German RSK (reactor safety commission), the corresponding French organization GPR and the German SSK (radiation protection commission). This publication continues earlier joint recommendations by the national bodies, last published by the German BMU (responsible German ministry) on 5 May 1995, in BAnz. page 7452. The safety recommendations establish the basis for further activities in the Franco-German project for development of the EPR, a PWR type reactor of the next generation. (CB) [de

  15. Genetics of leaf rust-resistant mutant WH 147-LM-1 in hexaploid wheat variety WH 147

    International Nuclear Information System (INIS)

    Reddy, V.R.K.; Viswanathan, P.

    1999-01-01

    By applying gamma rays, EMS and their combination in hexaploid wheat variety WH 147, a total of 20 mutants (0.0226%) exhibiting complete leaf rust resistance were isolated from segregating M2 rows.When one of the rust-resistant mutants, WH 147-LM-1 was crossed with the universally susceptible, suggesting that the mutant character is controlled by one dominant gene and one recessive gene.The F2 plants derived by crossing the mutant WH 147-LM with seven near-isogenic wheat lines showed segregation for susceptibility, indicating that the mutant character was indeed generated through induced mutations

  16. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  17. Replacement of the control and instrumentation system with the microprocessor based systems in Japanese PWR plants

    International Nuclear Information System (INIS)

    Hayashi, N.

    1998-01-01

    In Ohi Units 3 and 4, Ikata Unit 3, and Genkai Units 3 and 4, the latest of PWR plants now under operation in Japan, the reactor control system and turbine control system employ the microprocessor base digital control systems with a view to improving reliability, operability and maintainability. In the next stage plants, another application of such digital system is also planned for the instrumentation rack for the reactor protection system for further improvement. On the other hand, in Mihama Unit 1, the first of domestic PWR plants, and later plants except for the latest 5 plants, analog control systems are employed for the instrumentation racks. For the analog control systems of these plants, FOXBORO H-Line instruments, equivalent domestic box type instruments or WH7300 Series card type instruments were initially employed, and later replaced with domestic card type control systems after 10-15 year operation. However, 8-12 years have passed since these replacements, so the 15th year generally quoted as an interval for replacing C and I systems is near at hand. This is the time to consider next replacement. This replacement will be based on the latest digital technology. However, it is not practical way for the existing plants to apply the same integrated digital C and I system configuration for the next stage plants, because it requires the drastic change of the C and I system configuration and significant cost-up. Therefore, we must investigate the optimum digital C and I system configuration for the existing system. (author)

  18. 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

  19. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  20. 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

  1. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  2. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  3. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  5. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  6. 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.)

  7. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  8. Subject-Auxillary Inversion Errors and Wh Question Acquisition: What Children Do Know.

    Science.gov (United States)

    Rowland, Caroline F.; Pine, Julian M.

    2000-01-01

    Analyzed correct wh-question production and subject-auxiliary inversion errors in one child's wh-question data. Argues that two current movement rule accounts cannot explain patterning of early wh-questions. Data can be explained by the child's knowledge of particular lexically-specific wh-word+auxiliary combinations, and inversion and universion…

  9. tWH associated production at the LHC

    Energy Technology Data Exchange (ETDEWEB)

    Demartin, Federico; Maltoni, Fabio [Universite catholique de Louvain, Centre for Cosmology, Particle Physics and Phenomenology (CP3), Louvain-la-Neuve (Belgium); Maier, Benedikt [Karlsruher Institut fuer Technologie (KIT), Institut fuer Experimentelle Kernphysik, Karlsruhe (Germany); Massachusetts Institute of Technology, Laboratory for Nuclear Science, Cambridge, MA (United States); Mawatari, Kentarou [Universite Grenoble-Alpes, CNRS/IN2P3, Laboratoire de Physique Subatomique et de Cosmologie, Grenoble (France); Vrije Universiteit Brussel, Theoretische Natuurkunde and IIHE/ELEM, Brussels (Belgium); International Solvay Institutes, Brussels (Belgium); Zaro, Marco [Sorbonne Universites, UPMC Univ. Paris 06, UMR 7589, LPTHE, Paris (France); CNRS, UMR 7589, LPTHE, Paris (France)

    2017-01-15

    We study Higgs boson production in association with a top quark and a W boson at the LHC. At NLO in QCD, tWH interferes with t anti tH and a procedure to meaningfully separate the two processes needs to be employed. In order to define tWH production for both total rates and differential distributions, we consider the diagram removal and diagram subtraction techniques that have been previously proposed for treating intermediate resonances at NLO, in particular in the context of tW production. These techniques feature approximations that need to be carefully taken into account when theoretical predictions are compared to experimental measurements. To this aim, we first critically revisit the tW process, for which an extensive literature exists and where an analogous interference with t anti t production takes place. We then provide robust results for total and differential cross sections for tW and tWH at 13 TeV, also matching short-distance events to a parton shower. We formulate a reliable prescription to estimate the theoretical uncertainties, including those associated to the very definition of the process at NLO. Finally, we study the sensitivity to a non-Standard-Model relative phase between the Higgs couplings to the top quark and to the W boson in tWH production. (orig.)

  10. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    Energy Technology Data Exchange (ETDEWEB)

    Bakalov, Ivan [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Berlin (Germany); Sonnenkalb, Martin [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany)

    2018-02-15

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  11. Investigation of conditions inside the reactor building annulus of a PWR plant of KONVOI type in case of severe accidents with increased containment leakages

    International Nuclear Information System (INIS)

    Bakalov, Ivan; Sonnenkalb, Martin

    2018-01-01

    Improvements of the implemented severe accident management (SAM) concepts have been done in all operating German NPPs after the Fukushima Daiichi accidents following recommendations of the German Reactor Safety Commission (RSK) and as a result of the stress test being performed. The efficiency of newly developed severe accident management guidelines (SAMG) for a PWR KONVOI reference plant related to the mitigation of challenging conditions inside the reactor building (RB) annulus due to increased containment leakages during severe accidents have been assessed. Based on two representative severe accident scenarios the releases of both hydrogen and radionuclides into the RB annulus have been predicted with different boundary conditions. The accident scenarios have been analysed without and with the impact of several SAM measures (already planned or proposed in addition), which turned out to be efficient to mitigate the consequences. The work was done within the frame of a research project financially supported by the Federal Ministry BMUB.

  12. Bituminization of simulated PWR type reactor wastes, boric evaporator bottons and ion exchange resins, carried out in CNEN/SP using commercial bitumen available in the Brazilian market

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1986-01-01

    The first results of the study of bituminization of simulated PWR wastes, boric evaporator bottons and spent ion-exclange resins (OH - , H + ) and incinerated ash-wates are presented and discussed. The study consisted of characterization of the commercial bitumen, locally available and bitumen wastes products of varying whight compositions. The characterization was carried out using standard analysis methods of ABNT and ASTM, and included measurement of, penetration, softening point and flash point. In addition, the bitumen samples were analized for their resin and asphaltene contents. For leaching studies, wastes products of bitumen and resin loaded with 134 Cs were utilized. The method used was according to the ISO norms. The simulation of the industrial process was carried out using an extruder-evaporator typically used in the plastic industries offered by Industria de Maquinas Miotto Ltda., Sao Bernardo do Campo - SP. (Author) [pt

  13. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    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. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  14. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

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

  15. 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

  16. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    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) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  17. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    International Nuclear Information System (INIS)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago

    2017-01-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  18. Microstructural characterization of stainless steel 17-4 PH used in the control element of PWR-Type reactors submitted to different heat treatments

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Douglas F.A.; Rezende, Renato P.; Turcarelli, Tiago, E-mail: ferreira@marinha.mil.br, E-mail: renato.rezende@marinha.mil.br, E-mail: tiago.turcarelli@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (DDNM/CTMSP), São Paulo, SP (Brazil). Diretoria de Desenvolvimento Nuclear da Marinha

    2017-07-01

    The Control Element is a set of mechanical components of pressurized water cooled nuclear reactors (PWR), with the function of modifying the reactivity of the nucleus by insertion and withdrawal of neutron absorptive rod, in order to change the flow of neutrons (power) to the necessary and desired levels. The control element also has a safety function when there is a need to have negative reactivity available to shut down the reactor in normal operating or accident situations. In this situation, the control element descends instantly and inserts the rods with absorptive material into the fuel element thus shutting down the reactor. The control element consists of control rods, which carry the neutron absorption material and is supported by the spider, pin, spring and spring retainer assembly. The control element has some components that need to have high resistance to impacts when the safety function is activated, so the material of this component must have high mechanical strength and toughness. One of the materials in which can be specified for this application is martensitic stainless steel 17- 4PH (UNS 17400). This steel, when subjected to the aging heat treatment, has its mechanical properties altered due to the precipitation of dispersed intermetallic compounds in the matrix. In all heat treatments performed the predominant microstructure is lath martensite. The heat treatment of the 620 °C / 4 h presented lower hardness when compared to the other treatments and when increase time and temperature the material presents Nb precipitates that increase the hardness. (author)

  19. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  20. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  1. Additional functions of remotely read kWh-meters

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, P. [VTT Energy, Espoo (Finland); Vehvilaeinen, S. [Mittrix Oy (Finland); Rantanen, J. [Helsinki Energy Board, Helsinki (Finland)

    1996-12-31

    In this presentation the possibilities to include new applications into remotely read smart kWh-meters are considered. New electronic meters can measure various electric quantities and have some extra calculating capacity. So they can be used to provide functions that distribution automation and the customer need and thus share the costs. Some applications like monitoring the state of the distribution network or locating faults are only for the utility, but many applications also need an interface to the customer or his automation system. Among those are monitoring the quality of electricity, estimating load curves, applying dynamic tariffs and selling electricity and accounting. As a special item, the continuous monitoring of the quality of electricity is discussed. This includes voltage levels, total distortion, asymmetry and so on. If such a kWh-meter indicates quality problems the situation can be detected with a suitable portable quality meter. The poor quality can be detected before it causes harm to equipment owned by the customers or the power distribution company. This presentation also presents a prototype of such a quality monitoring kWh-meter

  2. Additional functions of remotely read kWh-meters

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, P [VTT Energy, Espoo (Finland); Vehvilaeinen, S [Mittrix Oy (Finland); Rantanen, J [Helsinki Energy Board, Helsinki (Finland)

    1997-12-31

    In this presentation the possibilities to include new applications into remotely read smart kWh-meters are considered. New electronic meters can measure various electric quantities and have some extra calculating capacity. So they can be used to provide functions that distribution automation and the customer need and thus share the costs. Some applications like monitoring the state of the distribution network or locating faults are only for the utility, but many applications also need an interface to the customer or his automation system. Among those are monitoring the quality of electricity, estimating load curves, applying dynamic tariffs and selling electricity and accounting. As a special item, the continuous monitoring of the quality of electricity is discussed. This includes voltage levels, total distortion, asymmetry and so on. If such a kWh-meter indicates quality problems the situation can be detected with a suitable portable quality meter. The poor quality can be detected before it causes harm to equipment owned by the customers or the power distribution company. This presentation also presents a prototype of such a quality monitoring kWh-meter

  3. 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)

  4. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  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. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  7. Orality in Osundare's Poetry: “WH-ASK” and “NP-WH-LET ...

    African Journals Online (AJOL)

    Two rhetorical structural types, derived, to be sure, from the Yoruba ... generic forms from Yoruba verbal habits, leaving out the content. ... must have influenced the poet. ..... family. Logba has however not shown any grammatical feature of locative .... (17), the verb kpé ['eat'] is marked with mo NEG which occurs before.

  8. The time course of processing difficulties with non-WH extraction in Danish

    DEFF Research Database (Denmark)

    Poulsen, Mads

    Danish, a V2-language, allows liberal extraction of non-WH elements from a variety of clause types to sentence-initial position, e.g. from relative clauses. Extractions from complement clauses, see (1), are more frequent (Jensen 2001) than extractions from adverbial clauses as in (2). (1)    De......-verbs), the parser doesn’t expect to have linked all mentioned arguments with the main verb at the clause boundary. For clauses with intransitive verbs, on the other hand, the parser expects to be able to find a role for all constituents within the clause, i.e. the parser should experience difficulties...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  10. 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)

  11. 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

  12. 10 CFR 431.402 - Preemption of State regulations for commercial HVAC & WH products.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 3 2010-01-01 2010-01-01 false Preemption of State regulations for commercial HVAC & WH... regulations for commercial HVAC & WH products. Beginning on the effective date of such standard, an energy conservation standard set forth in this Part for a commercial HVAC & WH product supersedes any State or local...

  13. Additional functions of remotely read kWh-meters

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, P [VTT Energy, Espoo (Finland); Vehvilaeinen, S [Mittrix Oy (Finland); Rantanen, J [Helsinki Energy Board (Finland)

    1998-08-01

    In this chapter the possibilities to include new applications into remotely read smart kWh-meters are considered. New electronic meters can measure various electric quantities and have some extra calculating capacity. So they can be used to provide functions that distribution automation and the customer need and thus share the costs. Some applications like monitoring the state of the distribution network or locating faults are only for the utility, but many applications also need an interface to the customers or their automation systems. Among those are monitoring the quality of electricity, estimating load curves, applying dynamic tariffs and selling electricity and accounting. As a special item, the continuous monitoring of the quality of electricity is discussed. This includes voltage levels, total distortion, asymmetry and so on. If such a kWh-meter indicates quality problems it is possible to go there with a portable quality meter that is suitable for the case and inspect the situation. The poor quality can be detected before it causes harm to equipment owned by the customers or the power distribution company. This article also presents a prototype of such a quality monitoring kWh-meter. Dynamic tariffs and free electricity markets require two way communication with the utility and the customer and measurement of the time variations of the energy consumption. The customers or their energy management system must receive the energy prices from the utility and calculate the energy costs and decide upon the energy consumption control actions. Some alternative ways to meet these customer interface requirements are compared. Remote reading of kWh-meters requires a certain investment in meters and their data communication with the utility. Because smart meters can have some additional memory and calculating capacity and are capable of measuring various electric quantities, it is possible to share the costs with other applications that use the same hardware and data

  14. 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

  15. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  16. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  17. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  18. 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

  19. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  20. 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.)

  1. Toxoplasma Chinese 1 Strain of WH3Δrop16I/III/gra15II Genetic Background Contributes to Abnormal Pregnant Outcomes in Murine Model

    Directory of Open Access Journals (Sweden)

    Cong Wang

    2018-06-01

    Full Text Available Toxoplasma gondii infection evokes a strong Th1-type response with interleukin (IL-12 and interferon (IFN-γ secretion. Recent studies suggest that the infection of pregnant mice with T. gondii may lead to adverse pregnancy results caused by subversion of physiological immune tolerance at maternofetal interface rather than direct invasion of the parasite. Genotype-associated dense granule protein GRA15II tends to induce classically activated macrophage (M1 differentiation and subsequently activating NK, Th1, and Th17 cells whereas rhoptry protein ROP16I/III drives macrophages to alternatively activated macrophage (M2 polarization and elicits Th2 immune response. Unlike the archetypal strains of types I, II, and III, type Chinese 1 strains possess both GRA15II and ROP16I/III, suggesting a distinct pathogenesis of Toxoplasma-involved adverse pregnancies. We constructed T. gondii type Chinese 1 strain of WH3Δrop16 based on CRISPR/Cas9 technology to explore the ROP16I/III-deficient/GRA15II-dominant parasites in induction of trophoblast apoptosis in vitro and abnormal pregnant outcomes of mice in vivo. Our study showed that Toxoplasma WH3Δrop16 remarkably induced apoptosis of trophoblasts. C57BL/6 pregnant mice injected with the tachyzoites of WH3Δrop16 presented increased absorptivity of fetuses in comparison with the mice infected with WH3 wild type (WH3 WT parasites although no remarkable difference of virulence to mice was seen between the two strains. Additionally, the mice inoculated with WH3Δrop16 tachyzoites exhibited a notable expression of both IL-17A and IFN-γ, while the percentage of CD4+CD25+FoxP3 [T regulatory cells (Tregs] were diminished in splenocytes and placenta tissues compared to those infected with WH3 WT parasites. Accordingly, expressions of IL-4, IL-10, and transforming growth factor beta 1, the pivotal cytokines of Th2 and Tregs response, were significantly dampened whereas IFN-γ and IL-12 expressions were

  2. The Auditory Comprehension of Wh-Questions in Aphasia: Support for the Intervener Hypothesis.

    Science.gov (United States)

    Sheppard, Shannon M; Walenski, Matthew; Love, Tracy; Shapiro, Lewis P

    2015-06-01

    This study examines 3 hypotheses about the processing of wh-questions in both neurologically healthy adults and adults with Broca's aphasia. We used an eye tracking while listening method with 32 unimpaired participants (Experiment 1) and 8 participants with Broca's aphasia (Experiment 2). Accuracy, response time, and online gaze data were collected. In Experiment 1, we established a baseline for how unimpaired processing and comprehension of 4 types of wh-question (subject- and object-extracted who- and which-questions) manifest. There was no unambiguous support found for any of the 3 hypotheses in Experiment 1. In Experiment 2 with the Broca's participants, however, we found significantly lower accuracy, slower response times, and increased interference in our gaze data in the object-extracted which-questions relative to the other conditions. Our results provide support for the intervener hypothesis, which states that sentence constructions that contain an intervener (a lexical noun phrase) between a displaced noun phrase and its gap site result in a significant processing disadvantage relative to other constructions. We argue that this hypothesis offers a compelling explanation for the comprehension deficits seen in some participants with Broca's aphasia.

  3. The improved damping of superconductor bearings for 35 kWh superconductor flywheel energy storage system

    International Nuclear Information System (INIS)

    Han, Y.H.; Park, B.J.; Jung, S.Y.; Han, S.C.; Lee, W.R.; Bae, Y.C.

    2013-01-01

    Highlights: ► We made a 35 kWh superconductor flywheel energy storage system. ► The damping coefficient of the superconductor bearing was increased over 3000 N s/m. ► The source of damping was discussed. -- Abstract: A 35 kWh Superconductor Flywheel Energy Storage system (SFES) using hybrid bearing sets, which is composed of a high temperature superconductor (HTS) bearing and an active magnet damper (AMD), has been developed at KEPCO Research Institute (KEPRI). Damping is a source of energy loss but necessary for the stability of the flywheel system. We found that the damping of HTS bearings can be improved by thermal insulating bolts, which play a role of passive type external damper. To investigate the source of the increased damping, damping coefficients were measured with HTS bearings using insulating bolts made of three kinds of polymer materials. The damping coefficient was raised over 3000 N s/m in the case of PEEK bolts. The value was almost a quarter of the AMD. In this study, thermoelastic and Coulomb friction damping mechanisms are discussed. The main damping mechanism was the thermoelastic damping of the bolts themselves. And interfacial gap between the insulating bolt and metal chamber, which increased during the cooling process, was considered to be the cause of the anisotropic damping coefficients. Finally, the effects of the HTS bearings on the first critical speed are shown

  4. 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.)

  5. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

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

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

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

  8. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Pitts, M.L.

    2000-01-01

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

  9. Search for massive WH resonances decaying into the $\\ell \

    CERN Document Server

    Khachatryan, Vardan; Tumasyan, Armen; Adam, Wolfgang; Aşılar, Ece; Bergauer, Thomas; Brandstetter, Johannes; Brondolin, Erica; Dragicevic, Marko; Erö, Janos; Flechl, Martin; Friedl, Markus; Fruehwirth, Rudolf; Ghete, Vasile Mihai; Hartl, Christian; Hörmann, Natascha; Hrubec, Josef; Jeitler, Manfred; Knünz, Valentin; König, Axel; Krammer, Manfred; Krätschmer, Ilse; Liko, Dietrich; Matsushita, Takashi; Mikulec, Ivan; Rabady, Dinyar; Rahbaran, Babak; Rohringer, Herbert; Schieck, Jochen; Schöfbeck, Robert; Strauss, Josef; Treberer-Treberspurg, Wolfgang; Waltenberger, Wolfgang; Wulz, Claudia-Elisabeth; Mossolov, Vladimir; Shumeiko, Nikolai; Suarez Gonzalez, Juan; Alderweireldt, Sara; Cornelis, Tom; De Wolf, Eddi A; Janssen, Xavier; Knutsson, Albert; Lauwers, Jasper; Luyckx, Sten; Van De Klundert, Merijn; Van Haevermaet, Hans; Van Mechelen, Pierre; Van Remortel, Nick; Van Spilbeeck, Alex; Abu Zeid, Shimaa; Blekman, Freya; D'Hondt, Jorgen; Daci, Nadir; De Bruyn, Isabelle; Deroover, Kevin; Heracleous, Natalie; Keaveney, James; Lowette, Steven; Moreels, Lieselotte; Olbrechts, Annik; Python, Quentin; Strom, Derek; Tavernier, Stefaan; Van Doninck, Walter; Van Mulders, Petra; Van Onsem, Gerrit Patrick; Van Parijs, Isis; Barria, Patrizia; Brun, Hugues; Caillol, Cécile; Clerbaux, Barbara; De Lentdecker, Gilles; Fasanella, Giuseppe; Favart, Laurent; Grebenyuk, Anastasia; Karapostoli, Georgia; Lenzi, Thomas; Léonard, Alexandre; Maerschalk, Thierry; Marinov, Andrey; Perniè, Luca; Randle-conde, Aidan; Reis, Thomas; Seva, Tomislav; Vander Velde, Catherine; Vanlaer, Pascal; Yonamine, Ryo; Zenoni, Florian; Zhang, Fengwangdong; Beernaert, Kelly; Benucci, Leonardo; Cimmino, Anna; Crucy, Shannon; Dobur, Didar; Fagot, Alexis; Garcia, Guillaume; Gul, Muhammad; Mccartin, Joseph; Ocampo Rios, Alberto Andres; Poyraz, Deniz; Ryckbosch, Dirk; Salva Diblen, Sinem; Sigamani, Michael; Tytgat, Michael; Van Driessche, Ward; Yazgan, Efe; Zaganidis, Nicolas; Basegmez, Suzan; Beluffi, Camille; Bondu, Olivier; Brochet, Sébastien; Bruno, Giacomo; Caudron, Adrien; Ceard, Ludivine; Da Silveira, Gustavo Gil; Delaere, Christophe; Favart, Denis; Forthomme, Laurent; Giammanco, Andrea; Hollar, Jonathan; Jafari, Abideh; Jez, Pavel; Komm, Matthias; Lemaitre, Vincent; Mertens, Alexandre; Musich, Marco; Nuttens, Claude; Perrini, Lucia; Pin, Arnaud; Piotrzkowski, Krzysztof; Popov, Andrey; Quertenmont, Loic; Selvaggi, Michele; Vidal Marono, Miguel; Beliy, Nikita; Hammad, Gregory Habib; Aldá Júnior, Walter Luiz; Alves, Fábio Lúcio; Alves, Gilvan; Brito, Lucas; Correa Martins Junior, Marcos; Hamer, Matthias; Hensel, Carsten; Mora Herrera, Clemencia; Moraes, Arthur; Pol, Maria Elena; Rebello Teles, Patricia; Belchior Batista Das Chagas, Ewerton; Carvalho, Wagner; Chinellato, Jose; Custódio, Analu; Melo Da Costa, Eliza; De Jesus Damiao, Dilson; De Oliveira Martins, Carley; Fonseca De Souza, Sandro; Huertas Guativa, Lina Milena; Malbouisson, Helena; Matos Figueiredo, Diego; Mundim, Luiz; Nogima, Helio; Prado Da Silva, Wanda Lucia; Santoro, Alberto; Sznajder, Andre; Tonelli Manganote, Edmilson José; Vilela Pereira, Antonio; Ahuja, Sudha; Bernardes, Cesar Augusto; De Souza Santos, Angelo; Dogra, Sunil; Tomei, Thiago; De Moraes Gregores, Eduardo; Mercadante, Pedro G; Moon, Chang-Seong; Novaes, Sergio F; Padula, Sandra; Romero Abad, David; Ruiz Vargas, José Cupertino; Aleksandrov, Aleksandar; Hadjiiska, Roumyana; Iaydjiev, Plamen; Rodozov, Mircho; Stoykova, Stefka; Sultanov, Georgi; Vutova, Mariana; Dimitrov, Anton; Glushkov, Ivan; Litov, Leander; Pavlov, Borislav; Petkov, Peicho; Ahmad, Muhammad; Bian, Jian-Guo; Chen, Guo-Ming; Chen, He-Sheng; Chen, Mingshui; Cheng, Tongguang; Du, Ran; Jiang, Chun-Hua; Plestina, Roko; Romeo, Francesco; Shaheen, Sarmad Masood; Spiezia, Aniello; Tao, Junquan; Wang, Chunjie; Wang, Zheng; Zhang, Huaqiao; Asawatangtrakuldee, Chayanit; Ban, Yong; Li, Qiang; Liu, Shuai; Mao, Yajun; Qian, Si-Jin; Wang, Dayong; Wang, Mengmeng; Xu, Zijun; Avila, Carlos; Cabrera, Andrés; Chaparro Sierra, Luisa Fernanda; Florez, Carlos; Gomez, Juan Pablo; Gomez Moreno, Bernardo; Sanabria, Juan Carlos; Godinovic, Nikola; Lelas, Damir; Puljak, Ivica; Ribeiro Cipriano, Pedro M; Antunovic, Zeljko; Kovac, Marko; Brigljevic, Vuko; Kadija, Kreso; Luetic, Jelena; Micanovic, Sasa; Sudic, Lucija; Attikis, Alexandros; Mavromanolakis, Georgios; Mousa, Jehad; Nicolaou, Charalambos; Ptochos, Fotios; Razis, Panos A; Rykaczewski, Hans; Bodlak, Martin; Finger, Miroslav; Finger Jr, Michael; Assran, Yasser; Elgammal, Sherif; Ellithi Kamel, Ali; Mahmoud, Mohammed; Calpas, Betty; Kadastik, Mario; Murumaa, Marion; Raidal, Martti; Tiko, Andres; Veelken, Christian; Eerola, Paula; Pekkanen, Juska; Voutilainen, Mikko; Härkönen, Jaakko; Karimäki, Veikko; Kinnunen, Ritva; Lampén, Tapio; Lassila-Perini, Kati; Lehti, Sami; Lindén, Tomas; Luukka, Panja-Riina; Mäenpää, Teppo; Peltola, Timo; Tuominen, Eija; Tuominiemi, Jorma; Tuovinen, Esa; Wendland, Lauri; Talvitie, Joonas; Tuuva, Tuure; Besancon, Marc; Couderc, Fabrice; Dejardin, Marc; Denegri, Daniel; Fabbro, Bernard; Faure, Jean-Louis; Favaro, Carlotta; Ferri, Federico; Ganjour, Serguei; Givernaud, Alain; Gras, Philippe; Hamel de Monchenault, Gautier; Jarry, Patrick; Locci, Elizabeth; Machet, Martina; Malcles, Julie; Rander, John; Rosowsky, André; Titov, Maksym; Zghiche, Amina; Antropov, Iurii; Baffioni, Stephanie; Beaudette, Florian; Busson, Philippe; Cadamuro, Luca; Chapon, Emilien; Charlot, Claude; Dahms, Torsten; Davignon, Olivier; Filipovic, Nicolas; Florent, Alice; Granier de Cassagnac, Raphael; Lisniak, Stanislav; Mastrolorenzo, Luca; Miné, Philippe; Naranjo, Ivo Nicolas; Nguyen, Matthew; Ochando, Christophe; Ortona, Giacomo; Paganini, Pascal; Pigard, Philipp; Regnard, Simon; Salerno, Roberto; Sauvan, Jean-Baptiste; Sirois, Yves; Strebler, Thomas; Yilmaz, Yetkin; Zabi, Alexandre; Agram, Jean-Laurent; Andrea, Jeremy; Aubin, Alexandre; Bloch, Daniel; Brom, Jean-Marie; Buttignol, Michael; Chabert, Eric Christian; Chanon, Nicolas; Collard, Caroline; Conte, Eric; Coubez, Xavier; Fontaine, Jean-Charles; Gelé, Denis; Goerlach, Ulrich; Goetzmann, Christophe; Le Bihan, Anne-Catherine; Merlin, Jeremie Alexandre; Skovpen, Kirill; Van Hove, Pierre; Gadrat, Sébastien; Beauceron, Stephanie; Bernet, Colin; Boudoul, Gaelle; Bouvier, Elvire; Carrillo Montoya, Camilo Andres; Chierici, Roberto; Contardo, Didier; Courbon, Benoit; Depasse, Pierre; El Mamouni, Houmani; Fan, Jiawei; Fay, Jean; Gascon, Susan; Gouzevitch, Maxime; Ille, Bernard; Lagarde, Francois; Laktineh, Imad Baptiste; Lethuillier, Morgan; Mirabito, Laurent; Pequegnot, Anne-Laure; Perries, Stephane; Ruiz Alvarez, José David; Sabes, David; Sgandurra, Louis; Sordini, Viola; Vander Donckt, Muriel; Verdier, Patrice; Viret, Sébastien; Toriashvili, Tengizi; Rurua, Lali; Autermann, Christian; Beranek, Sarah; Edelhoff, Matthias; Feld, Lutz; Heister, Arno; Kiesel, Maximilian Knut; Klein, Katja; Lipinski, Martin; Ostapchuk, Andrey; Preuten, Marius; Raupach, Frank; Schael, Stefan; Schulte, Jan-Frederik; Verlage, Tobias; Weber, Hendrik; Wittmer, Bruno; Zhukov, Valery; Ata, Metin; Brodski, Michael; Dietz-Laursonn, Erik; Duchardt, Deborah; Endres, Matthias; Erdmann, Martin; Erdweg, Sören; Esch, Thomas; Fischer, Robert; Güth, Andreas; Hebbeker, Thomas; Heidemann, Carsten; Hoepfner, Kerstin; Knutzen, Simon; Kreuzer, Peter; Merschmeyer, Markus; Meyer, Arnd; Millet, Philipp; Olschewski, Mark; Padeken, Klaas; Papacz, Paul; Pook, Tobias; Radziej, Markus; Reithler, Hans; Rieger, Marcel; Scheuch, Florian; Sonnenschein, Lars; Teyssier, Daniel; Thüer, Sebastian; Cherepanov, Vladimir; Erdogan, Yusuf; Flügge, Günter; Geenen, Heiko; Geisler, Matthias; Hoehle, Felix; Kargoll, Bastian; Kress, Thomas; Kuessel, Yvonne; Künsken, Andreas; Lingemann, Joschka; Nehrkorn, Alexander; Nowack, Andreas; Nugent, Ian Michael; Pistone, Claudia; Pooth, Oliver; Stahl, Achim; Aldaya Martin, Maria; Asin, Ivan; Bartosik, Nazar; Behnke, Olaf; Behrens, Ulf; Bell, Alan James; Borras, Kerstin; Burgmeier, Armin; Campbell, Alan; Choudhury, Somnath; Costanza, Francesco; Diez Pardos, Carmen; Dolinska, Ganna; Dooling, Samantha; Dorland, Tyler; Eckerlin, Guenter; Eckstein, Doris; Eichhorn, Thomas; Flucke, Gero; Gallo, Elisabetta; Garay Garcia, Jasone; Geiser, Achim; Gizhko, Andrii; Gunnellini, Paolo; Hauk, Johannes; Hempel, Maria; Jung, Hannes; Kalogeropoulos, Alexis; Karacheban, Olena; Kasemann, Matthias; Katsas, Panagiotis; Kieseler, Jan; Kleinwort, Claus; Korol, Ievgen; Lange, Wolfgang; Leonard, Jessica; Lipka, Katerina; Lobanov, Artur; Lohmann, Wolfgang; Mankel, Rainer; Marfin, Ihar; Melzer-Pellmann, Isabell-Alissandra; Meyer, Andreas Bernhard; Mittag, Gregor; Mnich, Joachim; Mussgiller, Andreas; Naumann-Emme, Sebastian; Nayak, Aruna; Ntomari, Eleni; Perrey, Hanno; Pitzl, Daniel; Placakyte, Ringaile; Raspereza, Alexei; Roland, Benoit; Sahin, Mehmet Özgür; Saxena, Pooja; Schoerner-Sadenius, Thomas; Schröder, Matthias; Seitz, Claudia; Spannagel, Simon; Trippkewitz, Karim Damun; Walsh, Roberval; Wissing, Christoph; Blobel, Volker; Centis Vignali, Matteo; Draeger, Arne-Rasmus; Erfle, Joachim; Garutti, Erika; Goebel, Kristin; Gonzalez, Daniel; Görner, Martin; Haller, Johannes; Hoffmann, Malte; Höing, Rebekka Sophie; Junkes, Alexandra; Klanner, Robert; Kogler, Roman; Kovalchuk, Nataliia; Lapsien, Tobias; Lenz, Teresa; Marchesini, Ivan; Marconi, Daniele; Meyer, Mareike; Nowatschin, Dominik; Ott, Jochen; Pantaleo, Felice; Peiffer, Thomas; Perieanu, Adrian; Pietsch, Niklas; Poehlsen, Jennifer; Rathjens, Denis; Sander, Christian; Scharf, Christian; Schettler, Hannes; Schleper, Peter; Schlieckau, Eike; Schmidt, Alexander; Schwandt, Joern; Sola, Valentina; Stadie, Hartmut; Steinbrück, Georg; Tholen, Heiner; Troendle, Daniel; Usai, Emanuele; Vanelderen, Lukas; Vanhoefer, Annika; Vormwald, Benedikt; Akbiyik, Melike; Barth, Christian; Baus, Colin; Berger, Joram; Böser, Christian; Butz, Erik; Chwalek, Thorsten; Colombo, Fabio; De Boer, Wim; Descroix, Alexis; Dierlamm, Alexander; Fink, Simon; Frensch, Felix; Friese, Raphael; Giffels, Manuel; Gilbert, Andrew; Haitz, Dominik; Hartmann, Frank; Heindl, Stefan Michael; Husemann, Ulrich; Katkov, Igor; Kornmayer, Andreas; Lobelle Pardo, Patricia; Maier, Benedikt; Mildner, Hannes; Mozer, Matthias Ulrich; Müller, Thomas; Müller, Thomas; Plagge, Michael; Quast, Gunter; Rabbertz, Klaus; Röcker, Steffen; Roscher, Frank; Sieber, Georg; Simonis, Hans-Jürgen; Stober, Fred-Markus Helmut; Ulrich, Ralf; Wagner-Kuhr, Jeannine; Wayand, Stefan; Weber, Marc; Weiler, Thomas; Wöhrmann, Clemens; Wolf, Roger; Anagnostou, Georgios; Daskalakis, Georgios; Geralis, Theodoros; Giakoumopoulou, Viktoria Athina; Kyriakis, Aristotelis; Loukas, Demetrios; Psallidas, Andreas; Topsis-Giotis, Iasonas; Agapitos, Antonis; Kesisoglou, Stilianos; Panagiotou, Apostolos; Saoulidou, Niki; Tziaferi, Eirini; Evangelou, Ioannis; Flouris, Giannis; Foudas, Costas; Kokkas, Panagiotis; Loukas, Nikitas; Manthos, Nikolaos; Papadopoulos, Ioannis; Paradas, Evangelos; Strologas, John; Bencze, Gyorgy; Hajdu, Csaba; Hazi, Andras; Hidas, Pàl; Horvath, Dezso; Sikler, Ferenc; Veszpremi, Viktor; Vesztergombi, Gyorgy; Zsigmond, Anna Julia; Beni, Noemi; Czellar, Sandor; Karancsi, János; Molnar, Jozsef; Szillasi, Zoltan; Bartók, Márton; Makovec, Alajos; Raics, Peter; Trocsanyi, Zoltan Laszlo; Ujvari, Balazs; Mal, Prolay; Mandal, Koushik; Sahoo, Deepak Kumar; Sahoo, Niladribihari; Swain, Sanjay Kumar; Bansal, Sunil; Beri, Suman Bala; Bhatnagar, Vipin; Chawla, Ridhi; Gupta, Ruchi; Bhawandeep, Bhawandeep; Kalsi, Amandeep Kaur; Kaur, Anterpreet; Kaur, Manjit; Kumar, Ramandeep; Mehta, Ankita; Mittal, Monika; Singh, Jasbir; Walia, Genius; Kumar, Ashok; Bhardwaj, Ashutosh; Choudhary, Brajesh C; Garg, Rocky Bala; Kumar, Ajay; Malhotra, Shivali; Naimuddin, Md; Nishu, Nishu; Ranjan, Kirti; Sharma, Ramkrishna; Sharma, Varun; Bhattacharya, Satyaki; Chatterjee, Kalyanmoy; Dey, Sourav; Dutta, Suchandra; Jain, Sandhya; Majumdar, Nayana; Modak, Atanu; Mondal, Kuntal; Mukherjee, Swagata; Mukhopadhyay, Supratik; Roy, Ashim; Roy, Debarati; Roy Chowdhury, Suvankar; Sarkar, Subir; Sharan, Manoj; Abdulsalam, Abdulla; Chudasama, Ruchi; Dutta, Dipanwita; Jha, Vishwajeet; Kumar, Vineet; Mohanty, Ajit Kumar; Pant, Lalit Mohan; Shukla, Prashant; Topkar, Anita; Aziz, Tariq; Banerjee, Sudeshna; Bhowmik, Sandeep; Chatterjee, Rajdeep Mohan; Dewanjee, Ram Krishna; Dugad, Shashikant; Ganguly, Sanmay; Ghosh, Saranya; Guchait, Monoranjan; Gurtu, Atul; Kole, Gouranga; Kumar, Sanjeev; Mahakud, Bibhuprasad; Maity, Manas; Majumder, Gobinda; Mazumdar, Kajari; Mitra, Soureek; Mohanty, Gagan Bihari; Parida, Bibhuti; Sarkar, Tanmay; Sur, Nairit; Sutar, Bajrang; Wickramage, Nadeesha; Chauhan, Shubhanshu; Dube, Sourabh; Kothekar, Kunal; Sharma, Seema; Bakhshiansohi, Hamed; Behnamian, Hadi; Etesami, Seyed Mohsen; Fahim, Ali; Goldouzian, Reza; Khakzad, Mohsen; Mohammadi Najafabadi, Mojtaba; Naseri, Mohsen; Paktinat Mehdiabadi, Saeid; Rezaei Hosseinabadi, Ferdos; Safarzadeh, Batool; Zeinali, Maryam; Felcini, Marta; Grunewald, Martin; Abbrescia, Marcello; Calabria, Cesare; Caputo, Claudio; Colaleo, Anna; Creanza, Donato; Cristella, Leonardo; De Filippis, Nicola; De Palma, Mauro; Fiore, Luigi; Iaselli, Giuseppe; Maggi, Giorgio; Maggi, Marcello; Miniello, Giorgia; My, Salvatore; Nuzzo, Salvatore; Pompili, Alexis; Pugliese, Gabriella; Radogna, Raffaella; Ranieri, Antonio; Selvaggi, Giovanna; Silvestris, Lucia; Venditti, Rosamaria; Verwilligen, Piet; Abbiendi, Giovanni; Battilana, Carlo; Benvenuti, Alberto; Bonacorsi, Daniele; Braibant-Giacomelli, Sylvie; Brigliadori, Luca; Campanini, Renato; Capiluppi, Paolo; Castro, Andrea; Cavallo, Francesca Romana; Chhibra, Simranjit Singh; Codispoti, Giuseppe; Cuffiani, Marco; Dallavalle, Gaetano-Marco; Fabbri, Fabrizio; Fanfani, Alessandra; Fasanella, Daniele; Giacomelli, Paolo; Grandi, Claudio; Guiducci, Luigi; Marcellini, Stefano; Masetti, Gianni; Montanari, Alessandro; Navarria, Francesco; Perrotta, Andrea; Rossi, Antonio; Rovelli, Tiziano; Siroli, Gian Piero; Tosi, Nicolò; Travaglini, Riccardo; Cappello, Gigi; Chiorboli, Massimiliano; Costa, Salvatore; Di Mattia, Alessandro; Giordano, Ferdinando; Potenza, Renato; Tricomi, Alessia; Tuve, Cristina; Barbagli, Giuseppe; Ciulli, Vitaliano; Civinini, Carlo; D'Alessandro, Raffaello; Focardi, Ettore; Gonzi, Sandro; Gori, Valentina; Lenzi, Piergiulio; Meschini, Marco; Paoletti, Simone; Sguazzoni, Giacomo; Tropiano, Antonio; Viliani, Lorenzo; Benussi, Luigi; Bianco, Stefano; Fabbri, Franco; Piccolo, Davide; Primavera, Federica; Calvelli, Valerio; Ferro, Fabrizio; Lo Vetere, Maurizio; Monge, Maria Roberta; Robutti, Enrico; Tosi, Silvano; Brianza, Luca; Dinardo, Mauro Emanuele; Fiorendi, Sara; Gennai, Simone; Gerosa, Raffaele; Ghezzi, Alessio; Govoni, Pietro; Malvezzi, Sandra; Manzoni, Riccardo Andrea; Marzocchi, Badder; Menasce, Dario; Moroni, Luigi; Paganoni, Marco; Pedrini, Daniele; Ragazzi, Stefano; Redaelli, Nicola; Tabarelli de Fatis, Tommaso; Buontempo, Salvatore; Cavallo, Nicola; Di Guida, Salvatore; Esposito, Marco; Fabozzi, Francesco; Iorio, Alberto Orso Maria; Lanza, Giuseppe; Lista, Luca; Meola, Sabino; Merola, Mario; Paolucci, Pierluigi; Sciacca, Crisostomo; Thyssen, Filip; Azzi, Patrizia; Bacchetta, Nicola; Benato, Lisa; Bisello, Dario; Boletti, Alessio; Branca, Antonio; Carlin, Roberto; Carvalho Antunes De Oliveira, Alexandra; Checchia, Paolo; Dall'Osso, Martino; Dorigo, Tommaso; Dosselli, Umberto; Gasparini, Fabrizio; Gasparini, Ugo; Gozzelino, Andrea; Kanishchev, Konstantin; Lacaprara, Stefano; Margoni, Martino; Meneguzzo, Anna Teresa; Pazzini, Jacopo; Pozzobon, Nicola; Ronchese, Paolo; Simonetto, Franco; Torassa, Ezio; Tosi, Mia; Zanetti, Marco; Zotto, Pierluigi; Zucchetta, Alberto; Zumerle, Gianni; Braghieri, Alessandro; Magnani, Alice; Montagna, Paolo; Ratti, Sergio P; Re, Valerio; Riccardi, Cristina; Salvini, Paola; Vai, Ilaria; Vitulo, Paolo; Alunni Solestizi, Luisa; Biasini, Maurizio; Bilei, Gian Mario; Ciangottini, Diego; Fanò, Livio; Lariccia, Paolo; Mantovani, Giancarlo; Menichelli, Mauro; Saha, Anirban; Santocchia, Attilio; Androsov, Konstantin; Azzurri, Paolo; Bagliesi, Giuseppe; Bernardini, Jacopo; Boccali, Tommaso; Castaldi, Rino; Ciocci, Maria Agnese; Dell'Orso, Roberto; Donato, Silvio; Fedi, Giacomo; Foà, Lorenzo; Giassi, Alessandro; Grippo, Maria Teresa; Ligabue, Franco; Lomtadze, Teimuraz; Martini, Luca; Messineo, Alberto; Palla, Fabrizio; Rizzi, Andrea; Savoy-Navarro, Aurore; Serban, Alin Titus; Spagnolo, Paolo; Tenchini, Roberto; Tonelli, Guido; Venturi, Andrea; Verdini, Piero Giorgio; Barone, Luciano; Cavallari, Francesca; D'imperio, Giulia; Del Re, Daniele; Diemoz, Marcella; Gelli, Simone; Jorda, Clara; Longo, Egidio; Margaroli, Fabrizio; Meridiani, Paolo; Organtini, Giovanni; Paramatti, Riccardo; Preiato, Federico; Rahatlou, Shahram; Rovelli, Chiara; Santanastasio, Francesco; Traczyk, Piotr; Amapane, Nicola; Arcidiacono, Roberta; Argiro, Stefano; Arneodo, Michele; Bellan, Riccardo; Biino, Cristina; Cartiglia, Nicolo; Costa, Marco; Covarelli, Roberto; Degano, Alessandro; Demaria, Natale; Finco, Linda; Kiani, Bilal; Mariotti, Chiara; Maselli, Silvia; Migliore, Ernesto; Monaco, Vincenzo; Monteil, Ennio; Obertino, Maria Margherita; Pacher, Luca; Pastrone, Nadia; Pelliccioni, Mario; Pinna Angioni, Gian Luca; Ravera, Fabio; Romero, Alessandra; Ruspa, Marta; Sacchi, Roberto; Solano, Ada; Staiano, Amedeo; Tamponi, Umberto; Belforte, Stefano; Candelise, Vieri; Casarsa, Massimo; Cossutti, Fabio; Della Ricca, Giuseppe; Gobbo, Benigno; La Licata, Chiara; Marone, Matteo; Schizzi, Andrea; Zanetti, Anna; Kropivnitskaya, Anna; Nam, Soon-Kwon; Kim, Dong Hee; Kim, Gui Nyun; Kim, Min Suk; Kong, Dae Jung; Lee, Sangeun; Oh, Young Do; Sakharov, Alexandre; Son, Dong-Chul; Brochero Cifuentes, Javier Andres; Kim, Hyunsoo; Kim, Tae Jeong; Song, Sanghyeon; Choi, Suyong; Go, Yeonju; Gyun, Dooyeon; Hong, Byung-Sik; Jo, Mihee; Kim, Hyunchul; Kim, Yongsun; Lee, Byounghoon; Lee, Kisoo; Lee, Kyong Sei; Lee, Songkyo; Park, Sung Keun; Roh, Youn; Yoo, Hwi Dong; Choi, Minkyoo; Kim, Hyunyong; Kim, Ji Hyun; Lee, Jason Sang Hun; Park, Inkyu; Ryu, Geonmo; Ryu, Min Sang; Choi, Young-Il; Goh, Junghwan; Kim, Donghyun; Kwon, Eunhyang; Lee, Jongseok; Yu, Intae; Dudenas, Vytautas; Juodagalvis, Andrius; Vaitkus, Juozas; Ahmed, Ijaz; Ibrahim, Zainol Abidin; Komaragiri, Jyothsna Rani; Md Ali, Mohd Adli Bin; Mohamad Idris, Faridah; Wan Abdullah, Wan Ahmad Tajuddin; Yusli, Mohd Nizam; Casimiro Linares, Edgar; Castilla-Valdez, Heriberto; De La Cruz-Burelo, Eduard; Heredia-De La Cruz, Ivan; Hernandez-Almada, Alberto; Lopez-Fernandez, Ricardo; Sánchez Hernández, Alberto; Carrillo Moreno, Salvador; Vazquez Valencia, Fabiola; Pedraza, Isabel; Salazar Ibarguen, Humberto Antonio; Morelos Pineda, Antonio; Krofcheck, David; Butler, Philip H; Ahmad, Ashfaq; Ahmad, Muhammad; Hassan, Qamar; Hoorani, Hafeez R; Khan, Wajid Ali; Khurshid, Taimoor; Shoaib, Muhammad; Bialkowska, Helena; Bluj, Michal; Boimska, Bożena; Frueboes, Tomasz; Górski, Maciej; Kazana, Malgorzata; Nawrocki, Krzysztof; Romanowska-Rybinska, Katarzyna; Szleper, Michal; Zalewski, Piotr; Brona, Grzegorz; Bunkowski, Karol; Byszuk, Adrian; Doroba, Krzysztof; Kalinowski, Artur; Konecki, Marcin; Krolikowski, Jan; Misiura, Maciej; Olszewski, Michal; Walczak, Marek; Bargassa, Pedrame; Beirão Da Cruz E Silva, Cristóvão; Di Francesco, Agostino; Faccioli, Pietro; Ferreira Parracho, Pedro Guilherme; Gallinaro, Michele; Leonardo, Nuno; Lloret Iglesias, Lara; Nguyen, Federico; Rodrigues Antunes, Joao; Seixas, Joao; Toldaiev, Oleksii; Vadruccio, Daniele; Varela, Joao; Vischia, Pietro; Afanasiev, Serguei; Bunin, Pavel; Gavrilenko, Mikhail; Golutvin, Igor; Gorbunov, Ilya; Kamenev, Alexey; Karjavin, Vladimir; Konoplyanikov, Viktor; Lanev, Alexander; Malakhov, Alexander; Matveev, Viktor; Moisenz, Petr; Palichik, Vladimir; Perelygin, Victor; Shmatov, Sergey; Shulha, Siarhei; Skatchkov, Nikolai; Smirnov, Vitaly; Zarubin, Anatoli; Golovtsov, Victor; Ivanov, Yury; Kim, Victor; Kuznetsova, Ekaterina; Levchenko, Petr; Murzin, Victor; Oreshkin, Vadim; Smirnov, Igor; Sulimov, Valentin; Uvarov, Lev; Vavilov, Sergey; Vorobyev, Alexey; Andreev, Yuri; Dermenev, Alexander; Gninenko, Sergei; Golubev, Nikolai; Karneyeu, Anton; Kirsanov, Mikhail; Krasnikov, Nikolai; Pashenkov, Anatoli; Tlisov, Danila; Toropin, Alexander; Epshteyn, Vladimir; Gavrilov, Vladimir; Lychkovskaya, Natalia; Popov, Vladimir; Pozdnyakov, Ivan; Safronov, Grigory; Spiridonov, Alexander; Vlasov, Evgueni; Zhokin, Alexander; Bylinkin, Alexander; Andreev, Vladimir; Azarkin, Maksim; Dremin, Igor; Kirakosyan, Martin; Leonidov, Andrey; Mesyats, Gennady; Rusakov, Sergey V; Baskakov, Alexey; Belyaev, Andrey; Boos, Edouard; Bunichev, Viacheslav; Dubinin, Mikhail; Dudko, Lev; Ershov, Alexander; Gribushin, Andrey; Klyukhin, Vyacheslav; Kodolova, Olga; Lokhtin, Igor; Miagkov, Igor; Obraztsov, Stepan; Savrin, Viktor; Snigirev, Alexander; Azhgirey, Igor; Bayshev, Igor; Bitioukov, Sergei; Kachanov, Vassili; Kalinin, Alexey; Konstantinov, Dmitri; Krychkine, Victor; Petrov, Vladimir; Ryutin, Roman; Sobol, Andrei; Tourtchanovitch, Leonid; Troshin, Sergey; Tyurin, Nikolay; Uzunian, Andrey; Volkov, Alexey; Adzic, Petar; Milosevic, Jovan; Rekovic, Vladimir; Alcaraz Maestre, Juan; Calvo, Enrique; Cerrada, Marcos; Chamizo Llatas, Maria; Colino, Nicanor; De La Cruz, Begona; Delgado Peris, Antonio; Domínguez Vázquez, Daniel; Escalante Del Valle, Alberto; Fernandez Bedoya, Cristina; Fernández Ramos, Juan Pablo; Flix, Jose; Fouz, Maria Cruz; Garcia-Abia, Pablo; Gonzalez Lopez, Oscar; Goy Lopez, Silvia; Hernandez, Jose M; Josa, Maria Isabel; Navarro De Martino, Eduardo; Pérez-Calero Yzquierdo, Antonio María; Puerta Pelayo, Jesus; Quintario Olmeda, Adrián; Redondo, Ignacio; Romero, Luciano; Santaolalla, Javier; Senghi Soares, Mara; Albajar, Carmen; de Trocóniz, Jorge F; Missiroli, Marino; Moran, Dermot; Cuevas, Javier; Fernandez Menendez, Javier; Folgueras, Santiago; Gonzalez Caballero, Isidro; Palencia Cortezon, Enrique; Vizan Garcia, Jesus Manuel; Cabrillo, Iban Jose; Calderon, Alicia; Castiñeiras De Saa, Juan Ramon; De Castro Manzano, Pablo; Duarte Campderros, Jordi; Fernandez, Marcos; Garcia-Ferrero, Juan; Gomez, Gervasio; Lopez Virto, Amparo; Marco, Jesus; Marco, Rafael; Martinez Rivero, Celso; Matorras, Francisco; Munoz Sanchez, Francisca Javiela; Piedra Gomez, Jonatan; Rodrigo, Teresa; Rodríguez-Marrero, Ana Yaiza; Ruiz-Jimeno, Alberto; Scodellaro, Luca; Trevisani, Nicolò; Vila, Ivan; Vilar Cortabitarte, Rocio; Abbaneo, Duccio; Auffray, Etiennette; Auzinger, Georg; Bachtis, Michail; Baillon, Paul; Ball, Austin; Barney, David; Benaglia, Andrea; Bendavid, Joshua; Benhabib, Lamia; Benitez, Jose F; Berruti, Gaia Maria; Bloch, Philippe; Bocci, Andrea; Bonato, Alessio; Botta, Cristina; Breuker, Horst; Camporesi, Tiziano; Castello, Roberto; Cerminara, Gianluca; D'Alfonso, Mariarosaria; D'Enterria, David; Dabrowski, Anne; Daponte, Vincenzo; David Tinoco Mendes, Andre; De Gruttola, Michele; De Guio, Federico; De Roeck, Albert; De Visscher, Simon; Di Marco, Emanuele; Dobson, Marc; Dordevic, Milos; Dorney, Brian; Du Pree, Tristan; Dünser, Marc; Dupont, Niels; Elliott-Peisert, Anna; Franzoni, Giovanni; Funk, Wolfgang; Gigi, Dominique; Gill, Karl; Giordano, Domenico; Girone, Maria; Glege, Frank; Guida, Roberto; Gundacker, Stefan; Guthoff, Moritz; Hammer, Josef; Harris, Philip; Hegeman, Jeroen; Innocente, Vincenzo; Janot, Patrick; Kirschenmann, Henning; Kortelainen, Matti J; Kousouris, Konstantinos; Krajczar, Krisztian; Lecoq, Paul; Lourenco, Carlos; Lucchini, Marco Toliman; Magini, Nicolo; Malgeri, Luca; Mannelli, Marcello; Martelli, Arabella; Masetti, Lorenzo; Meijers, Frans; Mersi, Stefano; Meschi, Emilio; Moortgat, Filip; Morovic, Srecko; Mulders, Martijn; Nemallapudi, Mythra Varun; Neugebauer, Hannes; Orfanelli, Styliani; Orsini, Luciano; Pape, Luc; Perez, Emmanuelle; Peruzzi, Marco; Petrilli, Achille; Petrucciani, Giovanni; Pfeiffer, Andreas; Piparo, Danilo; Racz, Attila; Rolandi, Gigi; Rovere, Marco; Ruan, Manqi; Sakulin, Hannes; Schäfer, Christoph; Schwick, Christoph; Seidel, Markus; Sharma, Archana; Silva, Pedro; Simon, Michal; Sphicas, Paraskevas; Steggemann, Jan; Stieger, Benjamin; Stoye, Markus; Takahashi, Yuta; Treille, Daniel; Triossi, Andrea; Tsirou, Andromachi; Veres, Gabor Istvan; Wardle, Nicholas; Wöhri, Hermine Katharina; Zagoździńska, Agnieszka; Zeuner, Wolfram Dietrich; Bertl, Willi; Deiters, Konrad; Erdmann, Wolfram; Horisberger, Roland; Ingram, Quentin; Kaestli, Hans-Christian; Kotlinski, Danek; Langenegger, Urs; Renker, Dieter; Rohe, Tilman; Bachmair, Felix; Bäni, Lukas; Bianchini, Lorenzo; Casal, Bruno; Dissertori, Günther; Dittmar, Michael; Donegà, Mauro; Eller, Philipp; Grab, Christoph; Heidegger, Constantin; Hits, Dmitry; Hoss, Jan; Kasieczka, Gregor; Lustermann, Werner; Mangano, Boris; Marionneau, Matthieu; Martinez Ruiz del Arbol, Pablo; Masciovecchio, Mario; Meister, Daniel; Micheli, Francesco; Musella, Pasquale; Nessi-Tedaldi, Francesca; Pandolfi, Francesco; Pata, Joosep; Pauss, Felicitas; Perrozzi, Luca; Quittnat, Milena; Rossini, Marco; Starodumov, Andrei; Takahashi, Maiko; Tavolaro, Vittorio Raoul; Theofilatos, Konstantinos; Wallny, Rainer; Aarrestad, Thea Klaeboe; Amsler, Claude; Caminada, Lea; Canelli, Maria Florencia; Chiochia, Vincenzo; De Cosa, Annapaola; Galloni, Camilla; Hinzmann, Andreas; Hreus, Tomas; Kilminster, Benjamin; Lange, Clemens; Ngadiuba, Jennifer; Pinna, Deborah; Robmann, Peter; Ronga, Frederic Jean; Salerno, Daniel; Yang, Yong; Cardaci, Marco; Chen, Kuan-Hsin; Doan, Thi Hien; Jain, Shilpi; Khurana, Raman; Konyushikhin, Maxim; Kuo, Chia-Ming; Lin, Willis; Lu, Yun-Ju; Yu, Shin-Shan; Kumar, Arun; Bartek, Rachel; Chang, Paoti; Chang, You-Hao; Chang, Yu-Wei; Chao, Yuan; Chen, Kai-Feng; Chen, Po-Hsun; Dietz, Charles; Fiori, Francesco; Grundler, Ulysses; Hou, George Wei-Shu; Hsiung, Yee; Liu, Yueh-Feng; Lu, Rong-Shyang; Miñano Moya, Mercedes; Petrakou, Eleni; Tsai, Jui-fa; Tzeng, Yeng-Ming; Asavapibhop, Burin; Kovitanggoon, Kittikul; Singh, Gurpreet; Srimanobhas, Norraphat; Suwonjandee, Narumon; Adiguzel, Aytul; Bakirci, Mustafa Numan; Cerci, Salim; Demiroglu, Zuhal Seyma; Dozen, Candan; Dumanoglu, Isa; Eskut, Eda; Girgis, Semiray; Gokbulut, Gul; Guler, Yalcin; Gurpinar, Emine; Hos, Ilknur; Kangal, Evrim Ersin; Kayis Topaksu, Aysel; Onengut, Gulsen; Ozdemir, Kadri; Polatoz, Ayse; Vergili, Mehmet; Zorbilmez, Caglar; Akin, Ilina Vasileva; Bilin, Bugra; Bilmis, Selcuk; Isildak, Bora; Karapinar, Guler; Yalvac, Metin; Zeyrek, Mehmet; Gülmez, Erhan; Kaya, Mithat; Kaya, Ozlem; Yetkin, Elif Asli; Yetkin, Taylan; Cakir, Altan; Cankocak, Kerem; Sen, Sercan; Vardarlı, Fuat Ilkehan; Grynyov, Boris; Levchuk, Leonid; Sorokin, Pavel; Aggleton, Robin; Ball, Fionn; Beck, Lana; Brooke, James John; Clement, Emyr; Cussans, David; Flacher, Henning; Goldstein, Joel; Grimes, Mark; Heath, Greg P; Heath, Helen F; Jacob, Jeson; Kreczko, Lukasz; Lucas, Chris; Meng, Zhaoxia; Newbold, Dave M; Paramesvaran, Sudarshan; Poll, Anthony; Sakuma, Tai; Seif El Nasr-storey, Sarah; Senkin, Sergey; Smith, Dominic; Smith, Vincent J; Bell, Ken W; Belyaev, Alexander; Brew, Christopher; Brown, Robert M; Calligaris, Luigi; Cieri, Davide; Cockerill, David JA; Coughlan, John A; Harder, Kristian; Harper, Sam; Olaiya, Emmanuel; Petyt, David; Shepherd-Themistocleous, Claire; Thea, Alessandro; Tomalin, Ian R; Williams, Thomas; Womersley, William John; Worm, Steven; Baber, Mark; Bainbridge, Robert; Buchmuller, Oliver; Bundock, Aaron; Burton, Darren; Casasso, Stefano; Citron, Matthew; Colling, David; Corpe, Louie; Cripps, Nicholas; Dauncey, Paul; Davies, Gavin; De Wit, Adinda; Della Negra, Michel; Dunne, Patrick; Elwood, Adam; Ferguson, William; Fulcher, Jonathan; Futyan, David; Hall, Geoffrey; Iles, Gregory; Kenzie, Matthew; Lane, Rebecca; Lucas, Robyn; Lyons, Louis; Magnan, Anne-Marie; Malik, Sarah; Nash, Jordan; Nikitenko, Alexander; Pela, Joao; Pesaresi, Mark; Petridis, Konstantinos; Raymond, David Mark; Richards, Alexander; Rose, Andrew; Seez, Christopher; Tapper, Alexander; Uchida, Kirika; Vazquez Acosta, Monica; Virdee, Tejinder; Zenz, Seth Conrad; Cole, Joanne; Hobson, Peter R; Khan, Akram; Kyberd, Paul; Leggat, Duncan; Leslie, Dawn; Reid, Ivan; Symonds, Philip; Teodorescu, Liliana; Turner, Mark; Borzou, Ahmad; Call, Kenneth; Dittmann, Jay; Hatakeyama, Kenichi; Liu, Hongxuan; Pastika, Nathaniel; Charaf, Otman; Cooper, Seth; Henderson, Conor; Rumerio, Paolo; Arcaro, Daniel; Avetisyan, Aram; Bose, Tulika; Fantasia, Cory; Gastler, Daniel; Lawson, Philip; Rankin, Dylan; Richardson, Clint; Rohlf, James; St John, Jason; Sulak, Lawrence; Zou, David; Alimena, Juliette; Berry, Edmund; Bhattacharya, Saptaparna; Cutts, David; Dhingra, Nitish; Ferapontov, Alexey; Garabedian, Alex; Hakala, John; Heintz, Ulrich; Laird, Edward; Landsberg, Greg; Mao, Zaixing; Narain, Meenakshi; Piperov, Stefan; Sagir, Sinan; Syarif, Rizki; Breedon, Richard; Breto, Guillermo; Calderon De La Barca Sanchez, Manuel; Chauhan, Sushil; Chertok, Maxwell; Conway, John; Conway, Rylan; Cox, Peter Timothy; Erbacher, Robin; Gardner, Michael; Ko, Winston; Lander, Richard; Mulhearn, Michael; Pellett, Dave; Pilot, Justin; Ricci-Tam, Francesca; Shalhout, Shalhout; Smith, John; Squires, Michael; Stolp, Dustin; Tripathi, Mani; Wilbur, Scott; Yohay, Rachel; Cousins, Robert; Everaerts, Pieter; Farrell, Chris; Hauser, Jay; Ignatenko, Mikhail; Saltzberg, David; Takasugi, Eric; Valuev, Vyacheslav; Weber, Matthias; Burt, Kira; Clare, Robert; Ellison, John Anthony; Gary, J William; Hanson, Gail; Heilman, Jesse; Paneva, Mirena Ivova; Jandir, Pawandeep; Kennedy, Elizabeth; Lacroix, Florent; Long, Owen Rosser; Luthra, Arun; Malberti, Martina; Olmedo Negrete, Manuel; Shrinivas, Amithabh; Wei, Hua; Wimpenny, Stephen; Yates, Brent; Branson, James G; Cerati, Giuseppe Benedetto; Cittolin, Sergio; D'Agnolo, Raffaele Tito; Derdzinski, Mark; Holzner, André; Kelley, Ryan; Klein, Daniel; Letts, James; Macneill, Ian; Olivito, Dominick; Padhi, Sanjay; Pieri, Marco; Sani, Matteo; Sharma, Vivek; Simon, Sean; Tadel, Matevz; Vartak, Adish; Wasserbaech, Steven; Welke, Charles; Würthwein, Frank; Yagil, Avraham; Zevi Della Porta, Giovanni; Bradmiller-Feld, John; Campagnari, Claudio; Dishaw, Adam; Dutta, Valentina; Flowers, Kristen; Franco Sevilla, Manuel; Geffert, Paul; George, Christopher; Golf, Frank; Gouskos, Loukas; Gran, Jason; Incandela, Joe; Mccoll, Nickolas; Mullin, Sam Daniel; Richman, Jeffrey; Stuart, David; Suarez, Indara; West, Christopher; Yoo, Jaehyeok; Anderson, Dustin; Apresyan, Artur; Bornheim, Adolf; Bunn, Julian; Chen, Yi; Duarte, Javier; Mott, Alexander; Newman, Harvey B; Pena, Cristian; Pierini, Maurizio; Spiropulu, Maria; Vlimant, Jean-Roch; Xie, Si; Zhu, Ren-Yuan; Andrews, Michael Benjamin; Azzolini, Virginia; Calamba, Aristotle; Carlson, Benjamin; Ferguson, Thomas; Paulini, Manfred; Russ, James; Sun, Menglei; Vogel, Helmut; Vorobiev, Igor; Cumalat, John Perry; Ford, William T; Gaz, Alessandro; Jensen, Frank; Johnson, Andrew; Krohn, Michael; Mulholland, Troy; Nauenberg, Uriel; Stenson, Kevin; Wagner, Stephen Robert; Alexander, James; Chatterjee, Avishek; Chaves, Jorge; Chu, Jennifer; Dittmer, Susan; Eggert, Nicholas; Mirman, Nathan; Nicolas Kaufman, Gala; Patterson, Juliet Ritchie; Rinkevicius, Aurelijus; Ryd, Anders; Skinnari, Louise; Soffi, Livia; Sun, Werner; Tan, Shao Min; Teo, Wee Don; Thom, Julia; Thompson, Joshua; Tucker, Jordan; Weng, Yao; Wittich, Peter; Abdullin, Salavat; Albrow, Michael; Anderson, Jacob; Apollinari, Giorgio; Banerjee, Sunanda; Bauerdick, Lothar AT; Beretvas, Andrew; Berryhill, Jeffrey; Bhat, Pushpalatha C; Bolla, Gino; Burkett, Kevin; Butler, Joel Nathan; Cheung, Harry; Chlebana, Frank; Cihangir, Selcuk; Elvira, Victor Daniel; Fisk, Ian; Freeman, Jim; Gottschalk, Erik; Gray, Lindsey; Green, Dan; Grünendahl, Stefan; Gutsche, Oliver; Hanlon, Jim; Hare, Daryl; Harris, Robert M; Hasegawa, Satoshi; Hirschauer, James; Hu, Zhen; Jayatilaka, Bodhitha; Jindariani, Sergo; Johnson, Marvin; Joshi, Umesh; Jung, Andreas Werner; Klima, Boaz; Kreis, Benjamin; Kwan, Simon; Lammel, Stephan; Linacre, Jacob; Lincoln, Don; Lipton, Ron; Liu, Tiehui; Lopes De Sá, Rafael; Lykken, Joseph; Maeshima, Kaori; Marraffino, John Michael; Martinez Outschoorn, Verena Ingrid; Maruyama, Sho; Mason, David; McBride, Patricia; Merkel, Petra; Mishra, Kalanand; Mrenna, Stephen; Nahn, Steve; Newman-Holmes, Catherine; O'Dell, Vivian; Pedro, Kevin; Prokofyev, Oleg; Rakness, Gregory; Sexton-Kennedy, Elizabeth; Soha, Aron; Spalding, William J; Spiegel, Leonard; Strobbe, Nadja; Taylor, Lucas; Tkaczyk, Slawek; Tran, Nhan Viet; Uplegger, Lorenzo; Vaandering, Eric Wayne; Vernieri, Caterina; Verzocchi, Marco; Vidal, Richard; Weber, Hannsjoerg Artur; Whitbeck, Andrew; Yang, Fan; Acosta, Darin; Avery, Paul; Bortignon, Pierluigi; Bourilkov, Dimitri; Carnes, Andrew; Carver, Matthew; Curry, David; Das, Souvik; Field, Richard D; Furic, Ivan-Kresimir; Gleyzer, Sergei V; Hugon, Justin; Konigsberg, Jacobo; Korytov, Andrey; Low, Jia Fu; Ma, Peisen; Matchev, Konstantin; Mei, Hualin; Milenovic, Predrag; Mitselmakher, Guenakh; Rank, Douglas; Rossin, Roberto; Shchutska, Lesya; Snowball, Matthew; Sperka, David; Terentyev, Nikolay; Thomas, Laurent; Wang, Jian; Wang, Sean-Jiun; Yelton, John; Hewamanage, Samantha; Linn, Stephan; Markowitz, Pete; Martinez, German; Rodriguez, Jorge Luis; Ackert, Andrew; Adams, Jordon Rowe; Adams, Todd; Askew, Andrew; Bochenek, Joseph; Diamond, Brendan; Haas, Jeff; Hagopian, Sharon; Hagopian, Vasken; Johnson, Kurtis F; Khatiwada, Ajeeta; Prosper, Harrison; Weinberg, Marc; Baarmand, Marc M; Bhopatkar, Vallary; Colafranceschi, Stefano; Hohlmann, Marcus; Kalakhety, Himali; Noonan, Daniel; Roy, Titas; Yumiceva, Francisco; Adams, Mark Raymond; Apanasevich, Leonard; Berry, Douglas; Betts, Russell Richard; Bucinskaite, Inga; Cavanaugh, Richard; Evdokimov, Olga; Gauthier, Lucie; Gerber, Cecilia Elena; Hofman, David Jonathan; Kurt, Pelin; O'Brien, Christine; Sandoval Gonzalez, Irving Daniel; Silkworth, Christopher; Turner, Paul; Varelas, Nikos; Wu, Zhenbin; Zakaria, Mohammed; Bilki, Burak; Clarida, Warren; Dilsiz, Kamuran; Durgut, Süleyman; Gandrajula, Reddy Pratap; Haytmyradov, Maksat; Khristenko, Viktor; Merlo, Jean-Pierre; Mermerkaya, Hamit; Mestvirishvili, Alexi; Moeller, Anthony; Nachtman, Jane; Ogul, Hasan; Onel, Yasar; Ozok, Ferhat; Penzo, Aldo; Snyder, Christina; Tiras, Emrah; Wetzel, James; Yi, Kai; Anderson, Ian; Barnett, Bruce Arnold; Blumenfeld, Barry; Eminizer, Nicholas; Fehling, David; Feng, Lei; Gritsan, Andrei; Maksimovic, Petar; Martin, Christopher; Osherson, Marc; Roskes, Jeffrey; Cocoros, Alice; Sarica, Ulascan; Swartz, Morris; Xiao, Meng; Xin, Yongjie; You, Can; Baringer, Philip; Bean, Alice; Benelli, Gabriele; Bruner, Christopher; Kenny III, Raymond Patrick; Majumder, Devdatta; Malek, Magdalena; Murray, Michael; Sanders, Stephen; Stringer, Robert; Wang, Quan; Ivanov, Andrew; Kaadze, Ketino; Khalil, Sadia; Makouski, Mikhail; Maravin, Yurii; Mohammadi, Abdollah; Saini, Lovedeep Kaur; Skhirtladze, Nikoloz; Toda, Sachiko; Lange, David; Rebassoo, Finn; Wright, Douglas; Anelli, Christopher; Baden, Drew; Baron, Owen; Belloni, Alberto; Calvert, Brian; Eno, Sarah Catherine; Ferraioli, Charles; Gomez, Jaime; Hadley, Nicholas John; Jabeen, Shabnam; Kellogg, Richard G; Kolberg, Ted; Kunkle, Joshua; Lu, Ying; Mignerey, Alice; Shin, Young Ho; Skuja, Andris; Tonjes, Marguerite; Tonwar, Suresh C; Apyan, Aram; Barbieri, Richard; Baty, Austin; Bierwagen, Katharina; Brandt, Stephanie; Busza, Wit; Cali, Ivan Amos; Demiragli, Zeynep; Di Matteo, Leonardo; Gomez Ceballos, Guillelmo; Goncharov, Maxim; Gulhan, Doga; Iiyama, Yutaro; Innocenti, Gian Michele; Klute, Markus; Kovalskyi, Dmytro; Lai, Yue Shi; Lee, Yen-Jie; Levin, Andrew; Luckey, Paul David; Marini, Andrea Carlo; Mcginn, Christopher; Mironov, Camelia; Narayanan, Siddharth; Niu, Xinmei; Paus, Christoph; Ralph, Duncan; Roland, Christof; Roland, Gunther; Salfeld-Nebgen, Jakob; Stephans, George; Sumorok, Konstanty; Varma, Mukund; Velicanu, Dragos; Veverka, Jan; Wang, Jing; Wang, Ta-Wei; Wyslouch, Bolek; Yang, Mingming; Zhukova, Victoria; Dahmes, Bryan; Evans, Andrew; Finkel, Alexey; Gude, Alexander; Hansen, Peter; Kalafut, Sean; Kao, Shih-Chuan; Klapoetke, Kevin; Kubota, Yuichi; Lesko, Zachary; Mans, Jeremy; Nourbakhsh, Shervin; Ruckstuhl, Nicole; Rusack, Roger; Tambe, Norbert; Turkewitz, Jared; Acosta, John Gabriel; Oliveros, Sandra; Avdeeva, Ekaterina; Bloom, Kenneth; Bose, Suvadeep; Claes, Daniel R; Dominguez, Aaron; Fangmeier, Caleb; Gonzalez Suarez, Rebeca; Kamalieddin, Rami; Keller, Jason; Knowlton, Dan; Kravchenko, Ilya; Meier, Frank; Monroy, Jose; Ratnikov, Fedor; Siado, Joaquin Emilo; Snow, Gregory R; Alyari, Maral; Dolen, James; George, Jimin; Godshalk, Andrew; Harrington, Charles; Iashvili, Ia; Kaisen, Josh; Kharchilava, Avto; Kumar, Ashish; Rappoccio, Salvatore; Roozbahani, Bahareh; Alverson, George; Barberis, Emanuela; Baumgartel, Darin; Chasco, Matthew; Hortiangtham, Apichart; Massironi, Andrea; Morse, David Michael; Nash, David; Orimoto, Toyoko; Teixeira De Lima, Rafael; Trocino, Daniele; Wang, Ren-Jie; Wood, Darien; Zhang, Jinzhong; Hahn, Kristan Allan; Kubik, Andrew; Mucia, Nicholas; Odell, Nathaniel; Pollack, Brian; Pozdnyakov, Andrey; Schmitt, Michael Henry; Stoynev, Stoyan; Sung, Kevin; Trovato, Marco; Velasco, Mayda; Brinkerhoff, Andrew; Dev, Nabarun; Hildreth, Michael; Jessop, Colin; Karmgard, Daniel John; Kellams, Nathan; Lannon, Kevin; Lynch, Sean; Marinelli, Nancy; Meng, Fanbo; Mueller, Charles; Musienko, Yuri; Pearson, Tessa; Planer, Michael; Reinsvold, Allison; Ruchti, Randy; Smith, Geoffrey; Taroni, Silvia; Valls, Nil; Wayne, Mitchell; Wolf, Matthias; Woodard, Anna; Antonelli, Louis; Brinson, Jessica; Bylsma, Ben; Durkin, Lloyd Stanley; Flowers, Sean; Hart, Andrew; Hill, Christopher; Hughes, Richard; Ji, Weifeng; Kotov, Khristian; Ling, Ta-Yung; Liu, Bingxuan; Luo, Wuming; Puigh, Darren; Rodenburg, Marissa; Winer, Brian L; Wulsin, Howard Wells; Driga, Olga; Elmer, Peter; Hardenbrook, Joshua; Hebda, Philip; Koay, Sue Ann; Lujan, Paul; Marlow, Daniel; Medvedeva, Tatiana; Mooney, Michael; Olsen, James; Palmer, Christopher; Piroué, Pierre; Saka, Halil; Stickland, David; Tully, Christopher; Zuranski, Andrzej; Malik, Sudhir; Barnes, Virgil E; Benedetti, Daniele; Bortoletto, Daniela; Gutay, Laszlo; Jha, Manoj; Jones, Matthew; Jung, Kurt; Miller, David Harry; Neumeister, Norbert; Radburn-Smith, Benjamin Charles; Shi, Xin; Shipsey, Ian; Silvers, David; Sun, Jian; Svyatkovskiy, Alexey; Wang, Fuqiang; Xie, Wei; Xu, Lingshan; Parashar, Neeti; Stupak, John; Adair, Antony; Akgun, Bora; Chen, Zhenyu; Ecklund, Karl Matthew; Geurts, Frank JM; Guilbaud, Maxime; Li, Wei; Michlin, Benjamin; Northup, Michael; Padley, Brian Paul; Redjimi, Radia; Roberts, Jay; Rorie, Jamal; Tu, Zhoudunming; Zabel, James; Betchart, Burton; Bodek, Arie; de Barbaro, Pawel; Demina, Regina; Eshaq, Yossof; Ferbel, Thomas; Galanti, Mario; Garcia-Bellido, Aran; Han, Jiyeon; Harel, Amnon; Hindrichs, Otto; Khukhunaishvili, Aleko; Petrillo, Gianluca; Tan, Ping; Verzetti, Mauro; Arora, Sanjay; Barker, Anthony; Chou, John Paul; Contreras-Campana, Christian; Contreras-Campana, Emmanuel; Duggan, Daniel; Ferencek, Dinko; Gershtein, Yuri; Gray, Richard; Halkiadakis, Eva; Hidas, Dean; Hughes, Elliot; Kaplan, Steven; Kunnawalkam Elayavalli, Raghav; Lath, Amitabh; Nash, Kevin; Panwalkar, Shruti; Park, Michael; Salur, Sevil; Schnetzer, Steve; Sheffield, David; Somalwar, Sunil; Stone, Robert; Thomas, Scott; Thomassen, Peter; Walker, Matthew; Foerster, Mark; Riley, Grant; Rose, Keith; Spanier, Stefan; York, Andrew; Bouhali, Othmane; Castaneda Hernandez, Alfredo; Dalchenko, Mykhailo; De Mattia, Marco; Delgado, Andrea; Dildick, Sven; Eusebi, Ricardo; Gilmore, Jason; Kamon, Teruki; Krutelyov, Vyacheslav; Mueller, Ryan; Osipenkov, Ilya; Pakhotin, Yuriy; Patel, Rishi; Perloff, Alexx; Rose, Anthony; Safonov, Alexei; Tatarinov, Aysen; Ulmer, Keith; Akchurin, Nural; Cowden, Christopher; Damgov, Jordan; Dragoiu, Cosmin; Dudero, Phillip Russell; Faulkner, James; Kunori, Shuichi; Lamichhane, Kamal; Lee, Sung Won; Libeiro, Terence; Undleeb, Sonaina; Volobouev, Igor; Appelt, Eric; Delannoy, Andrés G; Greene, Senta; Gurrola, Alfredo; Janjam, Ravi; Johns, Willard; Maguire, Charles; Mao, Yaxian; Melo, Andrew; Ni, Hong; Sheldon, Paul; Snook, Benjamin; Tuo, Shengquan; Velkovska, Julia; Xu, Qiao; Arenton, Michael Wayne; Cox, Bradley; Francis, Brian; Goodell, Joseph; Hirosky, Robert; Ledovskoy, Alexander; Li, Hengne; Lin, Chuanzhe; Neu, Christopher; Sinthuprasith, Tutanon; Sun, Xin; Wang, Yanchu; Wolfe, Evan; Wood, John; Xia, Fan; Clarke, Christopher; Harr, Robert; Karchin, Paul Edmund; Kottachchi Kankanamge Don, Chamath; Lamichhane, Pramod; Sturdy, Jared; Belknap, Donald; Carlsmith, Duncan; Cepeda, Maria; Dasu, Sridhara; Dodd, Laura; Duric, Senka; Gomber, Bhawna; Grothe, Monika; Hall-Wilton, Richard; Herndon, Matthew; Hervé, Alain; Klabbers, Pamela; Lanaro, Armando; Levine, Aaron; Long, Kenneth; Loveless, Richard; Mohapatra, Ajit; Ojalvo, Isabel; Perry, Thomas; Pierro, Giuseppe Antonio; Polese, Giovanni; Ruggles, Tyler; Sarangi, Tapas; Savin, Alexander; Sharma, Archana; Smith, Nicholas; Smith, Wesley H; Taylor, Devin; Woods, Nathaniel

    2016-04-28

    A search for a massive resonance W' decaying into a W and a Higgs boson in the l nu b anti-b (l = e, mu) final state is presented. Results are based on data corresponding to an integrated luminosity of 19.7 inverse femtobarns of proton-proton collisions at sqrt(s)=8 TeV, collected using the CMS detector at the LHC. For a high-mass (>~1 TeV) resonance, the two bottom quarks coming from the Higgs boson decay are reconstructed as a single jet, which can be tagged by placing requirements on its substructure and flavour. Exclusion limits at 95% confidence level are set on the production cross section of a narrow resonance decaying into WH, as a function of its mass. In the context of a little Higgs model, a lower limit on the W' mass of 1.4 TeV is set. In a heavy vector triplet model that mimics the properties of composite Higgs models, a lower limit on the W' mass of 1.5 TeV is set. In the context of this model, the results are combined with related searches to obtain a lower limit on the W' mass of 1.8 TeV, the ...

  10. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  11. 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

  12. 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)

  13. 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

  14. Comparative accident analyses for a KONVOI-type PWR using the integral codes ASTEC V1.33 and MELCOR 1.8.6

    International Nuclear Information System (INIS)

    Reinke, Nils; Erdmann, Walter; Nowack, Holger; Sonnenkalb, Martin

    2010-08-01

    In the frame of the project RS1180 funded by the German Federal Ministry for Economics and Technology (BMWi) calculations have been carried out with the integral code ASTEC V1.33 p3 developed by GRS for two postulated accidents in a nuclear power plant with KONVOI type a pressurized water reactor and compared to calculations with MELCOR 1.8.6 YU. Major objective was to assess the capability of ASTEC for application in level 2 probabilistic safety analyses (PSA). In particular, it was investigated to which extent ASTEC is able to perform such integral calculations meeting criteria with regard to both reasonable calculation time and specific boundary conditions necessary for PSA analyses. Two exemplary accidents were selected: - A transient with loss of steam generator feed water, - A small break loss of coolant accident (50 cm 2 ) in the cold leg of the coolant line connected to the pressurizer. In principle, the results demonstrate the capability of ASTEC V1.33 to carry out such PSA level 2 calculations. In addition, it has to be noted that for both ASTEC and MELCOR the requirements in view of the quality of the results leads to prolonged calculation times due to more detailed nodalisations of the whole plant. This is valid for the core region as well as for the primary circuit and for the containment. Consequently, calculation times in the order of one day to two weeks are accomplished, thereby excluding extensive parameter analyses in order to assess the sensitivity of the calculation results. Concerning the quality of the results a good agreement can be stated between ASTEC and MELCOR results in terms of global data. In detail some results are sensitive to user effects. Here, the nodalisation seems to be of major influence besides differences in modeling specific phenomena. The comparison suggests that in particular the influence of the nodalisation defined by the user and depending on the user's experience should be carefully evaluated. Since some

  15. Error Patterns in Young German Children's "Wh"-Questions

    Science.gov (United States)

    Schmerse, Daniel; Lieven, Elena; Tomasello, Michael

    2013-01-01

    In this article we report two studies: a detailed longitudinal analysis of errors in "wh"-questions from six German-learning children (age 2 ; 0-3 ; 0) and an analysis of the prosodic characteristics of "wh"-questions in German child-directed speech. The results of the first study demonstrate that German-learning children…

  16. Technetium labeled WH701 for its potential use to image TNF-receptor-positive hepatocarcinoma

    International Nuclear Information System (INIS)

    Xia Jinsong; Wu Hua; Xiang Yan

    2004-01-01

    Objective: In this investigation, TNF analogs (WH701) was labeled with technetium (A number of TNF analogs had been selected and synthesized in our lab using random phage-display peptides library ) and pharmacokinetics and feasibility studies were performed for its potential use as diagnostic radiopharmaceutical. Methods WH701 was radiolabeled with 99m Tc then the complexes were characterized by thin layer chromatography. In vitro stability of the radiolabeled WH701 was examined simultaneity. Biodistribution and tumor uptake studies were also conducted to determine its in vivo characteristics. Results: The peptide analog WH701 permitted efficient incorporation of 99m Tc. The preparation of 99m Tc-WH701 was stable in vitro. Studies in vivo suggested that the biological activity of the peptide was not compromised. The agent was cleared rapidly from the blood and excreted mainly from kidney. The labeled peptide was shown in the nude mouse model to localize rapidly and specifically in site of tumor. Conclusions: The TNF analogue peptide WH701 can be radiolabeled with 99m Tc without loss of affinity, and the 99m Tc-WH701 shows radiochemical stability for an extended period of time in vitro. The high specific tumor uptake, rapid blood clearance, and predominantly renal excretion make 99m Tc-WH701 a promising candidate for tumor imaging. This agent is worthy of further investigation.

  17. Technetium labeled WH701 for its potential use to image TNF-receptor-positive hepatocarcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Jinsong, Xia; Hua, Wu; Yan, Xiang [Department of Nuclear Medicine, Tongji Hospital, Tongji Medical College, Huazhong University of Science and Technology, Wuhan (China)

    2004-07-01

    Objective: In this investigation, TNF analogs (WH701) was labeled with technetium (A number of TNF analogs had been selected and synthesized in our lab using random phage-display peptides library ) and pharmacokinetics and feasibility studies were performed for its potential use as diagnostic radiopharmaceutical. Methods WH701 was radiolabeled with {sup 99m}Tc then the complexes were characterized by thin layer chromatography. In vitro stability of the radiolabeled WH701 was examined simultaneity. Biodistribution and tumor uptake studies were also conducted to determine its in vivo characteristics. Results: The peptide analog WH701 permitted efficient incorporation of {sup 99m}Tc. The preparation of {sup 99m}Tc-WH701 was stable in vitro. Studies in vivo suggested that the biological activity of the peptide was not compromised. The agent was cleared rapidly from the blood and excreted mainly from kidney. The labeled peptide was shown in the nude mouse model to localize rapidly and specifically in site of tumor. Conclusions: The TNF analogue peptide WH701 can be radiolabeled with {sup 99m}Tc without loss of affinity, and the {sup 99m}Tc-WH701 shows radiochemical stability for an extended period of time in vitro. The high specific tumor uptake, rapid blood clearance, and predominantly renal excretion make {sup 99m}Tc-WH701 a promising candidate for tumor imaging. This agent is worthy of further investigation.

  18. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  19. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  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. The preparation of a radionuclide labeled peptide {sup 125}I-WH16 and its characters of binding to a human liver cancer cell line HepG2 in vitro

    Energy Technology Data Exchange (ETDEWEB)

    Sha, Luo; Xiaohua, Zhu; Hua, Wu [Department of Nuclear Medicine, Tongji Hospital, Tongji Medical College, Huazhong Univ. of Science and Technolgoy, Wuhan (China); Bing, Jia; Jing, Du; Fan, Wang

    2004-12-15

    Objective: To investigate the binding characters of a radionuclide labeled peptide {sup 125}I-WH16 which is affinitive to hepatocarcinoma cells in order to explore the potential feasibility of this artificially synthesized peptide to be a targeting reagent in diagnosis and therapy of liver cancer. Methods: 1) WH16 was labeled with Na{sup 125}I using Iodogen method, then isolated and identified with HPLC; 2)a. The tests of cell number (or time of incubation)- to-binding counts between {sup 125}I-WH16 and HepG2 cells were carried out in order to obtain better conditions for next in vitro tests; b. The average binding counts of {sup 125}I-WH16 treated HepG2 cells and L02 cells were compared in order to inspect the binding specificity between {sup 125}I-WH16 and HepG2 cells; c. A test of saturation of binding between {sup 125}I-WH16 and HepG2 cells was carried out in order to investigate the binding affinity between {sup 125}I-WH16 and HepG2 cells. Results: 1) The labeling rate of {sup 125}I was 50%. The specific activity of {sup 125}I-WH16 was 8.21x10{sup 15} Bq/mol. The radiochemical purity was 95% and the remnant radiochemical purity after a storage for 24 h at -20 degree C was 95%. The radioactive concentration was 6.64 x 10{sup 9} Bq/ L; 2) a. The binding of {sup 125}I-WH16 to HepG2 cells was cell number dependent and the optimal time of incubation was 3 h; b. There were obvious differences between HepG2 cells and L02 cells in binding with {sup 125}I-WH16; c. The binding of {sup 125}I-WH16 to HepG2 cells showed saturability. Scatchard plotting suggested that HepG2 cells contained only one type of WH16 receptors. The concentrations of Kd and Bmax were (1.42 {+-} 0.28) nmol/L and (12.15 {+-} 0.63) pmol/L, respectively. Hill modulus from Hill plotting was 1.03, which was close to 1 and suggesting that one receptor may bind only one ligand molecule. Conclusions: The present study indicates that the preparation of {sup 125}I-WH16 is stable and has good specificity and

  2. Steam generator for PWR type reactor

    International Nuclear Information System (INIS)

    Baba, Iwao; Hiyama, Nobuyuki.

    1994-01-01

    A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

  3. Abnormality diagnosis device for PWR type reactor

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The device of the present invention rapidly detects a small amount of primary coolants leaked from a heat transfer pipes in a steam generator to estimate the scale of the rupture and forecast the transition of the amount of the leakage. That is, a gamma-ray spectrum and dose rate measuring system having a bypass line capable of decaying short half-time nuclides disposed to each of second main steam pipelines and primary coolant pipelines. Data obtained from the measuring systems are compared, to judge the presence of abnormal symptom due to leakage and further, radioactivity concentration in main steams is calculated based on the dose rate at each of the measuring points. Further, radioactivity concentration of the leakage from the secondary main steam pipelines is calculated by change with passage of time. With such procedures, when primary coolants are leaked to the secondary main steam pipelines, the dose rate of the measured system is changed, thereby enabling to recognize the state at the abnormal point. The transition of abnormalities can be forecast with reference to the dose rate by change with passage of time. Further, state of the abnormal point can be recognized based on the gamma-ray spectrum, which is inherent upon occurrence of fuel rupture. (T.S.)

  4. Investigating the Grammatical and Pragmatic Origins of Wh-Questions in Children with Autism Spectrum Disorders

    Directory of Open Access Journals (Sweden)

    Manya Jyotishi

    2017-03-01

    Full Text Available Compared to typically developing children, children with autism (ASD show delayed production of wh-questions. It is currently controversial the degree to which such deficits derive from social-pragmatic requirements and/or because these are complex grammatical structures. The current study employed the intermodal preferential looking (IPL paradigm, which reduces social-pragmatic demands. The IPL paradigm can help distinguish these proposals, as successful comprehension promotes the “pragmatics-origins” argument whereas comprehension difficulties would implicate a “grammatical-origins” argument. Additionally, we tested both the linguistic and social explanations by assessing the contributions of children's early grammatical knowledge (i.e., SVO word order and their social-pragmatic scores on the Vineland to their later wh-question comprehension. Fourteen children with ASD and 17 TD children, matched on language level, were visited in their homes at 4-month intervals. Comprehension of wh-questions and SVO word order were tested via IPL: the wh-question video showed a costumed horse and bird serving as agents or patients of familiar transitive actions. During the test trials, they were displayed side by side with directing audios (e.g., “What did the horse tickle?”, “What hugged the bird?”, “Where is the horse/bird?”. Children's eye movements were coded offline; the DV was their percent looking to the named item during test. To show comprehension, children should look longer at the named item during a where-question than during a subject-wh or object-wh question. Results indicated that TD children comprehended both subject and object wh-questions at 32 months of age. Comprehension of object-wh questions emerged chronologically later in children with ASD compared to their TD peers, but at similar levels of language. Moreover, performance on word order and social-pragmatic scores independently predicted both groups' later

  5. Surveillance systems (PWR) - loose parts monitoring - vibration monitoring - leakage detection

    International Nuclear Information System (INIS)

    Schuette, A.; Blaesig, H.

    1982-01-01

    The contribution is engaged in the task and the results of the loose parts monitoring and the vibration monitoring following from the practice at the PWR of Biblis. First a description of both systems - location and type of the sensors used, the treatment of the measurements and the indications - is given. The results of the analysis of some events picked up by the surveillance systems are presented showing applicabilty and benefit of such systems. (orig.)

  6. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  7. PWR surveillance based on correspondence between empirical models and physical

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Upadhyaya, B.R.; Kerlin, T.W.

    1976-01-01

    An on line surveillance method based on the correspondence between empirical models and physicals models is proposed for pressurized water reactors. Two types of empirical models are considered as well as the mathematical models defining the correspondence between the physical and empirical parameters. The efficiency of this method is illustrated for the surveillance of the Doppler coefficient for Oconee I (an 886 MWe PWR) [fr

  8. Escape from the island. Processing constraints on wh-extraction in Danish

    DEFF Research Database (Denmark)

    Christensen, Ken Ramshøj; Kizach, Johannes; Nyvad, Anne Mette

    2012-01-01

    In the formal syntax literature, it is commonly assumed that there is a constraint on linguistic competence that blocks extraction of WH-expressions (e.g. what or which book) from embedded questions, referred to as WH-islands. Furthermore, it is assumed that there is an argument/adjunct asymmetry...... reveal that WH-island violations, though degraded, are grammatical in Danish. Since the standard assumptions cannot account for the range of results, we argue in favor of a processing account referring to Locality (processing domains) and Working Memory....

  9. 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)

  10. Influences of boric acid and lithium hydroxide on oxide film of type 316 stainless steel in PWR simulated primary water; PWR 1次冷却材模擬環境中の316ステンレス鋼に生成した皮膜性状に及ぼすほう酸および水酸化リチウムの影響

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Fukuya, Koji; Arioka, Koji [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan)

    2012-06-15

    In order to understand the influences of boric acid and lithium hydroxide on the IGSCC of type 316 stainless steel, an oxide film was analyzed in simulated PWR primary water while varying the boric acid and lithium hydroxide concentrations. It was found that, although boric acid and lithium hydroxide did not affect the structure and chemical composition of the surface oxide film remarkably, a lower boric acid concentration or a higher lithium concentration produced an oxide film with a thicker surface. It was considered that the lower boric acid concentration and higher lithium hydroxide concentration caused a higher magnetite solubility at the surface of the material and that the higher magnetite solubility caused a higher iron concentration gradient, which promoted iron dissolution from the material and the formation of a thicker oxide film. It was found that the thicker oxide film caused a higher IGSCC susceptibility and that the corrosion was the dominant factor of the IGSCC mechanism. No significant change was found in the morphologies of crack tip oxide in different bulk water chemistry systems, thus producing CT specimens with similar crack growth rates. (author)

  11. 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)

  12. 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)

  13. Performance of high burned PWR fuel during transient

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio

    1992-01-01

    In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

  14. 78 FR 15653 - Notice of Intent To Form the Commercial HVAC, WH, and Refrigeration Certification Working Group...

    Science.gov (United States)

    2013-03-12

    ... DEPARTMENT OF ENERGY 10 CFR Part 429 Notice of Intent To Form the Commercial HVAC, WH, and... Requirements for Commercial HVAC, WH, and Refrigeration Equipment AGENCY: Office of Energy Efficiency and... commercial heating, ventilation, and air-conditioning (HVAC), water heating (WH), and refrigeration equipment...

  15. Escape from the island. Grammaticality and (reduced) acceptability of wh-island violations in Danish

    DEFF Research Database (Denmark)

    Christensen, Ken Ramshøj; Kizach, Johannes; Nyvad, Anne Mette

    2013-01-01

    In the syntax literature, it is commonly assumed that a constraint on linguistic competence blocks extraction of wh-expressions (e.g. what or which book) from embedded questions, referred to as wh-islands. Furthermore, it is assumed that there is an argument/adjunct asymmetry in extraction from w...... the standard assumptions cannot account for the range of results, we argue in favor of a processing account referring to locality (processing domains) and working memory....

  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. 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

  18. Half a dollar per kWh and still cheap electricity

    International Nuclear Information System (INIS)

    Daey Ouwens, C.

    1991-01-01

    An analysis for the costs of electricity produced by an autonomous solar cell system in combination with efficient electric appliances, as they are used in houses and small offices, is presented. If efficient appliances are used, an average household will use about 700 kWh per year, compared to about 3000 kWh it is using on average at present. Efficient appliances, however, are more expensive. The extra cost per kWh saved is about the same as the cost of the fuel needed to produce a kWh in a power plant. The 700 kWh per year can be produced by an autonomous solar cell system with an area of 3 m 2 (efficiency 19%). The analysis is valid roughly for the area between 40 o northern and southern latitude. A comparison has been made between the total costs of a grid connection and the costs of an autonomous solar cell system. If all costs are capitalized and compiled, a grid connection will amount to about $6,000 (fuel costs are not included). The costs of a solar cell system producing 700 kWh per year are about the same or even lower. The modular nature of solar cell systems makes it possible to start with a certain number of appliances, depending on the amount of money one can spend. 6 tabs., 2 refs

  19. Quaternionic representation of the Coxeter group W(H4) and the polyhedra

    International Nuclear Information System (INIS)

    Koca, Mehmet; Al-Ajmi, Mudhahir; Koc, Ramazan

    2006-01-01

    The vertices of the four-dimensional polytope {3, 3, 5} and its dual {5, 3, 3} admitting the symmetry of the non-crystallographic Coxeter group W(H 4 ) of order 14,400 are represented in terms of quaternions with unit norm where the polytope {3, 3, 5} is represented by the elements of the binaryicosahedral group of quaternions of order 120. We projected the polytopes to three-dimensional Euclidean space where the quaternionic vertices are the orbits of the Coxeter group W(H 3 ), icosahedral group with inversion, where W(H 3 ) x Z 2 is one of the maximal subgroups of the Coxeter group W(H 4 ). The orbits of the icosahedral group W(H 3 ) in the polytope {3, 3, 5} are the conjugacy classes of the binary icosahedral group and represent a number of icosahedrons, dodecahedrons and one icosidodecahedron in three dimensions. The 15 orbits of the icosahedral group W(H 3 ) in the polytope {5, 3, 3} represent the dodecahedrons, icosidodecahedrons, small rhombicosidodecahedrons and some convex solids possessing the icosahedral symmetry. One of the convex solids with 60 vertices is very similar to the truncated icosahedron (soccer ball) but with two different edge lengths which can be taken as a realistic model of the C 60 molecule at extreme temperature and pressure

  20. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  1. 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)

  2. 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.

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  5. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  6. 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

  7. 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

  8. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  9. Syntactic constraints and individual differences in native and non-native processing of wh-movement

    Directory of Open Access Journals (Sweden)

    Adrienne eJohnson

    2016-04-01

    Full Text Available There is a debate as to whether second language (L2 learners show qualitatively similar processing profiles as native speakers or whether L2 learners are restricted in their ability to use syntactic information during online processing. In the realm of wh-dependency resolution, research has examined whether learners, similar to native speakers, attempt to resolve wh-dependencies in grammatically licensed contexts but avoid positing gaps in illicit contexts such as islands. Also at issue is whether the avoidance of gap filling in islands is due to adherence to syntactic constraints or whether islands simply present processing bottlenecks. One approach has been to examine the relationship between processing abilities and the establishment of wh-dependencies in islands. Grammatical accounts of islands do not predict such a relationship as the parser should simply not predict gaps in illicit contexts. In contrast, a pattern of results showing that individuals with more processing resources are better able to establish wh-dependencies in islands could conceivably be compatible with certain processing accounts. In a self-paced reading experiment which examines the processing of wh- dependencies, we address both questions, examining whether native English speakers and Korean learners of English show qualitatively similar patterns and whether there is a relationship between working memory, as measured by counting span and reading span, and processing in both island and non-island contexts. The results of the self-paced reading experiment suggest that learners can use syntactic information on the same timecourse as native speakers, showing qualitative similarity between the two groups. Results of regression analyses did not reveal a significant relationship between working memory and the establishment of wh-dependencies in islands but we did observe significant relationships between working memory and the processing of licit wh-dependencies. As the

  10. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

  11. Design and Construction of 10 kWh Class Flywheel Energy Storage System

    International Nuclear Information System (INIS)

    Jung, S. Y.; Han, S. C.; Han, Y. H.; Park, B. J.; Bae, Y. C.; Lee, W. R.

    2011-01-01

    A superconductor flywheel energy storage system (SFES) is an electro-mechanical battery which transforms electrical energy into mechanical energy for storage, and vice versa. A 10 kWh class flywheel energy storage system (FESS) has been developed to evaluate the feasibility of a 35 kWh class SFES with a flywheel Ip/If ratio larger than 1. The 10 kWh class FESS is composed of a main frame, a composite flywheel, active magnetic dampers (AMDs), a permanent magnet bearing, and a motor/generator. The flywheel of the FESS rotates at a very high speed to store energy, while being levitated by a permanent magnetic bearing and a pair of thrust AMDs. The 10 kWh class flywheel is mainly composed of a composite rotor assembly, where most of the energy is stored, two radial and two thrust AMD rotors, which dissipate vibration at critical speeds, a permanent magnet rotor, which supports most of the flywheel weight, a motor rotor, which spins the flywheel, and a central hollow shaft, where the parts are assembled and aligned to. The stators of each of the main components are assembled on to housings, which are assembled and aligned to the main frame. Many factors have been considered while designing each part of the flywheel, stator and frame. In this study, a 10 kWh class flywheel energy storage system has been designed and constructed for test operation.

  12. Wh-Questions, Universal Statements and Free Choice Inferences in Child Mandarin.

    Science.gov (United States)

    Huang, Haiquan; Zhou, Peng; Crain, Stephen

    2017-11-16

    This study investigated 5-year-old Mandarin-speaking children's comprehension of wh-questions, universal statements and free choice inferences. Previous research has found that Mandarin-speaking children assign a universal interpretation to sentences with a wh-word (e.g., shei 'who') followed by the adverbial quantifier dou 'all' (Zhou in Appl Psycholinguist 36:411-435, 2013). Children also compute free choice inferences in sentences that contain a modal verb in addition to a wh-word and dou (Zhou, in: Nakayama, Su, Huang (eds.) Studies in Chinese and Japanese language acquisition: in honour of Stephen Crain. John Benjamins Publishing Company, Amsterdam, pp 223-235, 2017). The present study used a Question-Statement Task to assess children's interpretation of sentences containing shei + dou, both with and without the modal verb beiyunxu 'was allowed to', as well as the contrast between sentences with shei + dou, which are statements for adults, versus ones with dou + shei, which are wh-questions for adults. The 5-year-old Mandarin-speaking child participants exhibited adult-like linguistic knowledge of the semantics and pragmatics of wh-words, the adverbial quantifier dou, and the deontic modal verb beiyunxu.

  13. 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

  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. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  16. 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

  17. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  18. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  19. 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.)

  20. The Impact of Similarity-Based Interference in Processing Wh-Questions in Aphasia

    Directory of Open Access Journals (Sweden)

    Shannon Mackenzie

    2014-04-01

    Full Text Available We describe two experiments investigating the comprehension of different types of Wh-questions in neurotypical adults (Experiment 1 and adults with Broca’s aphasia (Experiment 2. Consider as examples: Two mailmen and a fireman got into a fight yesterday afternoon. 1a. Who pushed the fireman yesterday afternoon? – Subject-extracted Who 1b. Who did the fireman push ___ yesterday afternoon? – Object-extracted Who 2a. Which mailman pushed the fireman yesterday afternoon? – Subject-extracted Which 2b. Which mailman did the fireman push ___ yesterday afternoon? – Object-extracted Which There is evidence from the linguistic and psycholinguistic literatures that suggest Which-questions are more difficult to understand than Who/What-questions and within those, that object-extracted are more difficult than subject-extracted. We used a unique eye tracking-while listening method where listeners were presented with sentences like (1 and (2 above while gazing at a three-figure picture (e.g., a picture of a mailman pushing a fireman who is pushing another mailman; we measured gazes to the referents in the pictures across the time-course of the sentences, and also collected accuracy and response time data to answer the questions (by button press. We examined four specific hypotheses: Discourse, Memory Retrieval, Word Order, and Intervener. The Discourse hypothesis suggests that Which-questions should be more difficult to process than Who-questions because the former is required to refer to an individual taken from a set of entities previously mentioned in the discourse (Donkers & Stowe, 2006; Shapiro, 2000. The Memory Retrieval hypothesis makes the opposite claim; Which-questions, unlike Who-questions, contain specific information in the Wh-phrase that should speed memory retrieval (Hofmeister, 2007. The Word Order hypothesis suggests that, regardless of question type (Which or Who, object-extracted questions should be more difficult to understand

  1. 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

  2. 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)

  3. 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)

    1998-12-31

    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)

  4. 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)

  5. Study of the noise propagation in PWR with coupled codes

    International Nuclear Information System (INIS)

    Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

    2011-01-01

    The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

  6. Underwater welding and repair technologies applied in PWR environment

    International Nuclear Information System (INIS)

    Scandella, Fabrice; Carpreau, Jean-Michel

    2012-01-01

    The authors describe several welding processes and technologies which have been used for underwater applications and which can be applied when repairing components of a PWR type reactor. They address, describe and discuss wet arc welding processes, the peculiarities of underwater welding, and the use of various processes such as 111, 114 and 135 processes, underwater welding with the hybrid plasma MIG-MAG process, underwater welding with the laser wire process, underwater welding with the FSW, FSP or UWFSW processes, underwater welding with variants of the friction welding process (friction surfacing, taper stitch welding, hydro-pillar processing

  7. 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

  8. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  9. Mathematical modelling of plant transients in the PWR for simulator purposes

    International Nuclear Information System (INIS)

    Hartel, K.

    1984-01-01

    This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

  10. 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

  11. 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.

  12. 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)

  13. Experimental results of a 3 k Wh thermochemical heat storage module for space heating application

    NARCIS (Netherlands)

    Finck, C.J.; Henquet, E.M.R.; Soest, C.F.L. van; Oversloot, H.P.; Jong, A.J. de; Cuypers, R.; Spijker, J.C. van 't

    2014-01-01

    A 3 kWh thermochemical heat storage (TCS) module was built as part of an all-in house system implementation focusing on space heating application at a temperature level of 40 ºC and a temperature lift of 20 K. It has been tested and measurements showed a maximum water circuit temperature span

  14. JBioWH: an open-source Java framework for bioinformatics data integration.

    Science.gov (United States)

    Vera, Roberto; Perez-Riverol, Yasset; Perez, Sonia; Ligeti, Balázs; Kertész-Farkas, Attila; Pongor, Sándor

    2013-01-01

    The Java BioWareHouse (JBioWH) project is an open-source platform-independent programming framework that allows a user to build his/her own integrated database from the most popular data sources. JBioWH can be used for intensive querying of multiple data sources and the creation of streamlined task-specific data sets on local PCs. JBioWH is based on a MySQL relational database scheme and includes JAVA API parser functions for retrieving data from 20 public databases (e.g. NCBI, KEGG, etc.). It also includes a client desktop application for (non-programmer) users to query data. In addition, JBioWH can be tailored for use in specific circumstances, including the handling of massive queries for high-throughput analyses or CPU intensive calculations. The framework is provided with complete documentation and application examples and it can be downloaded from the Project Web site at http://code.google.com/p/jbiowh. A MySQL server is available for demonstration purposes at hydrax.icgeb.trieste.it:3307. Database URL: http://code.google.com/p/jbiowh.

  15. Design of a 1-kWh bipolar nickel hydrogen battery

    Science.gov (United States)

    Cataldo, R. L.

    1984-01-01

    The design of a nickel hydrogen battery utilizing bipolar construction in a common pressure vessel is discussed. Design features are as follows: 40 ampere-hour capacity, 1 kWh stored energy as a 24 cell battery, 1.8 kW delivered in a LEO Cycle and maximum pulse power of 18.0 kW.

  16. Next-to-leading order QCD predictions for the hadronic WH+jet production

    International Nuclear Information System (INIS)

    Su Jijuan; Ma Wengan; Zhang Renyou; Guo Lei

    2010-01-01

    We calculate the next-to-leading order (NLO) QCD corrections to the WH 0 production in association with a jet at hadron colliders. We study the impacts of the complete NLO QCD radiative corrections to the integrated cross sections, the scale dependence of the cross sections, and the differential cross sections ((dσ/dcosθ), (dσ/dp T )) of the final W-, Higgs boson and jet. We find that the corrections significantly modify the physical observables, and reduce the scale uncertainty of the leading-order cross section. Our results show that by applying the inclusive scheme with p T,j cut =20 GeV and taking m H =120 GeV, μ=μ 0 ≡(1/2)(m W +m H ), the K-factor is 1.15 for the process pp→W ± H 0 j+X at the Tevatron, while the K-factors for the processes pp→W - H 0 j+X and pp→W + H 0 j+X at the LHC are 1.12 and 1.08, respectively. We conclude that to understand the hadronic associated WH 0 production, it is necessary to study the NLO QCD corrections to the WH 0 j production process which is part of the inclusive WH 0 production.

  17. The Auditory Comprehension of "Wh"-Questions in Aphasia: Support for the Intervener Hypothesis

    Science.gov (United States)

    Sheppard, Shannon M.; Walenski, Matthew; Love, Tracy; Shapiro, Lewis P.

    2015-01-01

    Purpose: This study examines 3 hypotheses about the processing of "wh"-questions in both neurologically healthy adults and adults with Broca's aphasia. Method: We used an eye tracking while listening method with 32 unimpaired participants (Experiment 1) and 8 participants with Broca's aphasia (Experiment 2). Accuracy, response time, and…

  18. Language and Variation: A Study of English and Persian Wh-questions

    Directory of Open Access Journals (Sweden)

    Laya Heidari Darani

    2015-05-01

    Full Text Available It was claimed by variationists that languages experience variation at all levels, which is supposed to be patterned. The present study aimed at exploring how variation occurred in English and Persian wh-questions. More specifically, it investigated whether such a variation was systematic and patterned. To this end, a modified version of the Edinburgh Map Task was used in data collection. The population of this study was 60 Canadian and Persian native speakers who performed the task through the construction of wh-question variants. The results indicated that both languages experienced variation in the construction of wh-questions. However, Persian proved to be more variable than English because it allowed more wh-question variants to be produced. Moreover, this variation was patterned in English and Persian individually, yet a systematic variation could not be observed between these two languages. It followed that the different mind-sets of the Canadian and Iranian participants which were affected by social and cultural factors might account for such a variation.

  19. Response to an indigenous smoking cessation media campaign - it's about whānau.

    Science.gov (United States)

    Grigg, Michele; Waa, Andrew; Bradbrook, Shane Kawenata

    2008-12-01

    To assess any effects among Māori (the indigenous people of New Zealand) smokers and their whānau (the traditional Māori family unit) of a campaign designed to support Māori smokers to quit smoking. New Zealand-wide cross sectional population surveys between 2000 and 2002 of smokers and whānau pre- and post-airing of the campaign. Measures included recall and awareness of the campaign; perceptions of the campaign; and campaign-attributed changes in quitting-related attitudes and behaviours. Seventy-eight per cent of smokers and 73% of whānau were able to recall the campaign one year following its launch. The television commercials (TVCs) were consistently rated very believable or very relevant by over half of the smokers who had seen them. More than half of smokers (54%) stated that the campaign had made them more likely to quit. This nationwide mass media cessation campaign developed to deliver a cessation message to indigenous people was received positively by Māori smokers and their whānau and played a role in prompting quit attempts. Social marketing campaigns have an important role as part of a tobacco control program to reduce high smoking prevalence among Māori and inequalities in health outcomes between Māori and other New Zealanders.

  20. Embedded "wh"-Questions in L2 English in India: Inversion as a Main Clause Phenomenon

    Science.gov (United States)

    Stringer, David

    2015-01-01

    This corpus study brings a second language (L2) research perspective, insights from generative grammar, and new empirical evidence to bear on a long-accepted claim in the World Englishes literature--namely, that inversion with "wh"-movement in colloquial Indian English is obligatory in embedded clauses and impossible in main clauses. It…

  1. LOFT CIS analysis 4''-WH-237-E inside containment penetration S-17B

    International Nuclear Information System (INIS)

    Nitzel, M.E.

    1978-01-01

    The stress analysis performed on the 4''-WH-237-E piping system inside containment penetration S-17B is presented. Deadweight, thermal expansion, and seismic loads were considered. Results of this analysis show that the subject piping system will meet ASME Code, Section III, Class 2 requirements

  2. JBioWH: an open-source Java framework for bioinformatics data integration

    Science.gov (United States)

    Vera, Roberto; Perez-Riverol, Yasset; Perez, Sonia; Ligeti, Balázs; Kertész-Farkas, Attila; Pongor, Sándor

    2013-01-01

    The Java BioWareHouse (JBioWH) project is an open-source platform-independent programming framework that allows a user to build his/her own integrated database from the most popular data sources. JBioWH can be used for intensive querying of multiple data sources and the creation of streamlined task-specific data sets on local PCs. JBioWH is based on a MySQL relational database scheme and includes JAVA API parser functions for retrieving data from 20 public databases (e.g. NCBI, KEGG, etc.). It also includes a client desktop application for (non-programmer) users to query data. In addition, JBioWH can be tailored for use in specific circumstances, including the handling of massive queries for high-throughput analyses or CPU intensive calculations. The framework is provided with complete documentation and application examples and it can be downloaded from the Project Web site at http://code.google.com/p/jbiowh. A MySQL server is available for demonstration purposes at hydrax.icgeb.trieste.it:3307. Database URL: http://code.google.com/p/jbiowh PMID:23846595

  3. Processing load imposed by line breaks in English temporal wh-questions

    Directory of Open Access Journals (Sweden)

    Masako Hirotani

    2016-10-01

    Full Text Available Prosody plays an important role in online sentence processing both explicitly and implicitly. It has been shown that prosodically packaging together parts of a sentence that are interpreted together facilitates processing of the sentence. This applies not only to explicit prosody but also implicit prosody. The present work hypothesizes that a line break in a written text induces an implicit prosodic break, which, in turn, should result in a processing bias for interpreting English wh-questions. Two experiments – one self-paced reading study and one questionnaire study – are reported. Both supported the line break hypothesis mentioned above. The results of the self-paced reading experiment showed that unambiguous wh-questions were read faster when the location of line breaks (or frame breaks matched the scope of a wh-phrase (main or embedded clause than when they did not. The questionnaire tested sentences with an ambiguous wh-phrase, one that could attach either to the main or the embedded clause. These sentences were interpreted as attaching to the main clause more often than to the embedded clause when a line break appeared after the main verb, but not when it appeared after the embedded verb.

  4. 29 CFR Appendix D to Part 825 - Notice of Eligibility and Rights & Responsibilities (Form WH-381)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 3 2010-07-01 2010-07-01 false Notice of Eligibility and Rights & Responsibilities (Form WH-381) D Appendix D to Part 825 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR OTHER LAWS THE FAMILY AND MEDICAL LEAVE ACT OF 1993 Pt. 825, App. D Appendix...

  5. Going beyond Input Quantity: "Wh"-Questions Matter for Toddlers' Language and Cognitive Development

    Science.gov (United States)

    Rowe, Meredith L.; Leech, Kathryn A.; Cabrera, Natasha

    2017-01-01

    There are clear associations between the overall quantity of input children are exposed to and their vocabulary acquisition. However, by uncovering specific features of the input that matter, we can better understand the mechanisms involved in vocabulary learning. We examine whether exposure to "wh"-questions, a challenging quality of…

  6. 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.

  7. 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.)

  8. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  9. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. Rupther: a simulation approach applied to a PWR vessel failure during a severe accident

    International Nuclear Information System (INIS)

    Mongabure, Ph.; Nicolas, L.; Devos, J.

    2000-01-01

    The Rupther program (Rupture Under Thermal Conditions) is a part of the international researches on severe accidents in the PWR type reactors. The aim of the program is the definition of failure simulation validated by experimental data on vessel steel 16MND5 mechanical properties. The paper presents the program and the first results. (A.L.B.)

  11. An investigation into the efficiency of ion-exchange membranes in simulated PWR coolants

    International Nuclear Information System (INIS)

    Clune, T.

    1980-11-01

    This report describes an investigation of the retention efficiency of cation-exchange membranes for magnesium, calcium and nickel ions in PWR-coolant type solutions containing 2 ppm lithium (as lithium hydroxide) and 1000 ppm boron (as boric acid). By analysis of the membranes themselves or of the effluent, the retention characteristics of the membranes in various experimental conditions have been examined. (author)

  12. Semi-automatic ultrasonic inspection of PWR upper internal immersed components

    International Nuclear Information System (INIS)

    Dombret, P.; Coquette, A.; Cermak, J.; Verspeelt, D.

    1985-01-01

    The present paper describes the characteristics of a semi-automatic ultrasonic inspection system. Components inspected are the so-called flexures, small pins located at the upper part of control rod tube-guide, some of which happened to broke in a few Westinghouse type PWR's. Inspection results and other system capabilities are also mentioned

  13. Response of pressurized water reactor (PWR) to network power generation demands

    International Nuclear Information System (INIS)

    Schreiner, L.A.

    1991-01-01

    The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

  14. Improved liquid waste processing system of PWR plant

    International Nuclear Information System (INIS)

    Suehiro, Kazuyasu

    1977-01-01

    Mitsubishi Heavy Industries, Ltd. has engaged in the improvement and enhancement of waste-processing facilities for PWR power stations, and recently established the improved processing system. With this system, it becomes possible to contain radioactive waste gas semi-permanently within plants and to recycle waste liquid after the treatment, thus to make the release of radioactive wastes practically zero. The improved system has the following features, namely the recycling system is adopted, drain is separated and each separated drain is treated by specialized process, the reboiler type evaporator and the reverse osmosis equipment are used, and the leakless construction is adopted for the equipments. The radioactive liquid wastes in PWR power stations are classified into coolant drain, drain from general equipments, chemical drain and cleaning water. The outline of the improved processing system and the newly developed equipments such as the reboiler type evaporator and the reverse osmosis equipment are explained. With the evaporator, the concentration rate of waste liquid can be raised to about three times, and foaming waste can be treated efficiently. The decontamination performance is excellent. The reverse osmosis treatment is stable and reliable method, and is useful for the treatment of cleaning water. It is also effective for concentrating treatment. The unmanned automatic operation is possible. (Kako, I.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F

    2003-12-01

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

  16. 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

  17. 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

  18. 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

  19. 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)

  20. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  1. 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)

  2. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  3. Mutations in actin used for structural studies partially disrupt β-thymosin/WH2 domains interaction.

    Science.gov (United States)

    Deville, Célia; Girard-Blanc, Christine; Assrir, Nadine; Nhiri, Naïma; Jacquet, Eric; Bontems, François; Renault, Louis; Petres, Stéphane; van Heijenoort, Carine

    2016-10-01

    Understanding the structural basis of actin cytoskeleton remodeling requires stabilization of actin monomers, oligomers, and filaments in complex with partner proteins, using various biochemical strategies. Here, we report a dramatic destabilization of the dynamic interaction with a model β-thymosin/WH2 domain induced by mutations in actin. This result underlines that mutant actins should be used with prudence to characterize interactions with intrinsically disordered partners as destabilization of dynamic interactions, although identifiable by NMR, may be invisible to other structural techniques. It also highlights how both β-thymosin/WH2 domains and actin tune local structure and dynamics in regulatory processes involving intrinsically disordered domains. © 2016 Federation of European Biochemical Societies.

  4. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  5. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  6. Development of MHI PWR fuel assembly with high thermal performance

    International Nuclear Information System (INIS)

    Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

  7. 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.

  8. 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

  9. An analysis of transients in the PWR downcomer

    International Nuclear Information System (INIS)

    Jovanovic, A.

    1981-01-01

    The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)

  10. 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)

  11. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  12. 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.)

  13. Interests and instrument: a micro-history of object Wh.3469 (X-ray powder diffraction camera, ca. 1940).

    Science.gov (United States)

    Scheffler, Robin Wolfe

    2009-12-01

    This paper presents a micro-history of an object in the collection of the Whipple Museum of the History of Science (accession no. Wh.3469), with an emphasis on how Wh.3469 reflects a hybrid of two different interwar British X-ray crystallographic communities, namely those based in WL Bragg's physics laboratory at the Victoria University of Manchester and the Crystallographic Laboratory at the University of Cambridge. It explores connections between Wh.3469's final design and construction and the different interests each community had in X-ray crystallography.

  14. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  15. 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

  16. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  17. 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

  18. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  19. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  20. 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

  1. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  2. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  3. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  4. 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)

  5. Shielding design for PWR in France

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  6. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  7. 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)

  8. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  9. 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

  10. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  11. 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.

  12. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  13. 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

  14. SCC of Alloy 600 components in PWR primary loop

    International Nuclear Information System (INIS)

    Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

    2002-01-01

    Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

  15. Evaluation of passive autocatalytic recombiners (PARS) performance for a PWR-konvoi containment type with Gothic 8.1 code; Evaluacion de la implementacion de recombinadores autocataliticos pasivos (PAR) en una contencion tipo Konvoi con el codigo Gothic 8.1

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Alonso Conty, E.; Papini, D.; Jimenez Varas, G.

    2016-08-01

    The study presented in this work analyses the evaluation of Passive Autocatalytic Recombiners (PARs) performance for a PWR-Konvoi containment type as a result of an international collaboration between the Paul Scherrer institute (PSI) and the Universidad Politecnica de Madrid (UPM). The implementation study analyzes the size, location and number of the PARs to minimize the risk arising from a hydrogen release and its distribution in the containment building during a hypothetical severe accident. A detailed 3D model of containment was used for the simulations developed for the Gothic 8.1 code. In the first place, the hydrogen preferential pathways and points of hydrogen accumulation were studies and identified starting from the base case scenario without any mitigation measure. The severe accident scenario chosen is a fast release of hydrogen-steam mixture from hot leg creep rupture during SBO (Station Black-Out) accident. Secondly a configuration of PARs was simulated under the same conditions of the unmitigated case. The PAR configuration offered an improvement in the chosen accident scenario, decreasing the hydrogen concentration values below the flammability limit /hydrogen concentration below 7%) in all the containment compartments. (Author)

  16. Fiscal implications of a 1-mill/kWh waste management fee

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-12-01

    The variations in parameters and uncertainties that can affect future waste management program costs are examined. Costs were estimated for both spent fuel and reprocessing waste disposal. Costs for repository site exploration and development, construction and operation of the first two repositories, and waste transportation of either spent fuel or reprocessing waste to the repository were included in the estimate of program costs. Costs for disposing of either spent fuel or reprocessing wastes in four generic geologic media (domed salt, bedded salt, tuff, and granite) were estimated considering uncertainties in package design waste preparation, mining cost, repository layout, repository startup date, and waste shipment distance. The range of these costs is then compared with estimated revenues from the proposed 1-mill/kWh fee to determine whether that fee would provide sufficient funds to meet waste management program needs. The estimated revenues and the range of program cost estimates in 1982 dollar resulting from these variations and uncertainties are shown graphically. The figure shows that the assumed 1-mill/kWh fee provides sufficient revenues to meet program costs for the range of conditions considered. Collection of a 1-mill/kWh fee at the time fuel is irradiated will result in the federal government precollecting for future waste management services. The relationship between funds accumulated to offset future waste management liabilities and the estimated cost of these liabilities will depend on the inflation rate and on the interest the federal government earns on the accumulated funds. If inflation continues, future fee adjustments will be required to assure that adequate funds are available to dispose of wastes when they are received

  17. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  18. Thermal behavior of an experimental 2.5-kWh lithium/iron sulfide battery

    Science.gov (United States)

    Chen, C. C.; Olszanski, T. W.; Gibbard, H. F.

    1981-10-01

    The thermal energy generation and the gross thermal energy balance in the battery systems was studied. High temperature lithium/iron sulfide batteries for electric vehicle applications were developed. The preferred battery temperature range during operation and idle periods is 400 to 500 C. Thermal management is an essential part of battery design, the battery requires a thermal insulation vessel to minimize heat loss and heating and cooling systems to control temperature. Results of temperature measurements performed on a 2.5-kWh battery module, which was built to gain information for the design of larger systems are reported.

  19. Tritium target performance during an LBLOCA in a PWR

    International Nuclear Information System (INIS)

    Reid, B.D.

    1996-01-01

    In December 1995, the U.S. Department of Energy (DOE) announced a preferred strategy for acquiring a new supply of tritium. That strategy is based on pursuing the two most promising production alternatives. These alternatives include either constructing an accelerator-produced tritium system for tritium production or procuring an existing commercial light water reactor or irradiation services from such a reactor to irradiate tritium targets. This paper discusses the safety performance of a tritium target in a commercial pressurized water reactor (PWR). The current conceptual design for the light water tritium targets is quite similar, in terms of external dimensions and materials, to early designs for stainless steel clad discrete burnable absorbers used in PWRs. The tritium targets nominally consist of an annular lithium aluminate pellet wrapped in a Zircaloy-4 getter and clad with Type 316 stainless steel

  20. 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.)

  1. Confirmation tests of PWR surveillance capsule shipping container

    International Nuclear Information System (INIS)

    Tomita, N.; Ue, K.; Ohashi, M.; Asada, K.; Yoneda, Y.

    1980-01-01

    Mitsubishi Heavy Industries, Ltd. carried out the confirmation tests to confirm the reliability of the PWR surveillance capsule shipping container and to collect cask design data using a 10-ton weight full scale model at Kobe Shipyard and Engine Works. This report presents the outline of these tests. The B Type container was a cylinder 3289 mm long, 1080 mm in diameter and designed in accordance with the new modified Japanese regulations similar to IAEA regulation. These tests consist of four 9 m drop tests, two 1 m puncture tests, a fire test and an immersion test. In conclusion, safetyness of this container has been proved and various technical data for cask design were also collected through these tests. (author)

  2. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

    The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

  3. 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

  4. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    Hobson, G.

    1984-01-01

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  5. 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

  6. Modeling local chemistry in PWR steam generator crevices

    International Nuclear Information System (INIS)

    Millett, P.J.

    1997-01-01

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

  7. Development of a lead acid battery suitable for electric vehicle propulsion. Final report. [96 V, 20 kWh, 50 Wh/kg

    Energy Technology Data Exchange (ETDEWEB)

    Schlotter, W J

    1977-08-26

    This report contains two detailed designs, and the design rationale, for an improved state-of-the-art electric vehicle battery incorporating expanded metal grids. The nominal 96-volt and 20-kWh battery incorporating this improved design is expected to cost about 25% less when manufactured in a mature plant. This report also contains detailed estimates for the capital cost and operating cost of a pilot plant to produce electric vehicle battery plates incorporating expanded metal grids. It is expected that the first electric vehicle batteries incorporating expanded metal grids can be available fifteen months after approval of this program. An additional program to improve lead acid batteries for electric vehicles further is also described. The advanced batteries resulting from this program are expected to incorporate either expanded metal grids and/or composite lead/plastic grids. In addition, these batteries are expected to contain low-density active materials. It is anticipated that those additional developments will result in an advanced battery capable of delivering 45 to 50 watt-hours/kg. As a result of the design and cost study, a ''First Buy'' improved state-of-the art vehicle battery proposed is included as part of this report. Eltra proposes to manufacture and deliver the required 2500 vehicle batteries within the time limits set forth by the Electric and Hybrid Vehicle Research, Development, and Demonstration Act of 1976. 20 figures, 13 tables.

  8. Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

  9. Steam turbines for PWR stations

    International Nuclear Information System (INIS)

    Muscroft, J.

    1989-01-01

    The thermodynamic cycle requirements and mechanical design features applying to modern GEC 3000 rev/min steam turbines for pressurised water reactor power stations are reviewed. The most recent developments include machines of 630 MW and 985 MW output which are currently under construction. The importance of service experience with nuclear wet steam turbines associated with a variety of types of water cooled reactor and its relevance to the design of modern 3000 rev/min turbines for pressurised water reactor applications is emphasised. (author)

  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. Development of Emergency Operating Strategies for Beyond Design Basis External Event(BDBEE)s in Korean WH Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hong, Soon-Joon; Hwang, Su-Hyun; Lee, Byung-Chul; Park, Kang-Min [FNC Tech. Co., Yongin (Korea, Republic of)

    2015-10-15

    Westinghouse developed and connected emergency operating procedures into a set of FLEX Support Guidelines(FSGs). This paper explains that Korean WH(Westinghouse) type nuclear power plants develop emergency operating strategies for ELAP(extended loss of all AC power), which include guidelines to use permanent and portable equipment as necessary to prevent core damage until AC power is restored from a reliable alternate source of AC power. The Korean emergency operating response strategies were developed to cope with a ELAP such as Fukushima event. The strategies include guidelines to prevent fuel damage using the FLEX equipment. Operators should take actions to prepare FLEX equipment within license basis SBO coping time. The loss of all AC power has been analyzed to identify the behavior of major NSSS process variables using RELAP computer code. The accident analysis showed that the plant does not result in fuel damage in 72 hours after an ELAP if operators take actions to cool RCS with opening of SG ADV in 5 gpm seal leak case. In this scenario, because ELAP is in process and all power cannot be used, operator should operate the FLEX equipment in order to actuate active equipment using the EOP fo SBO response. This strategy will prevent entering SAMG because this actions result in core cooling and stay in core exit temperature less than 650 .deg. C. Korean emergency operating guidelines(EOGs) will be developed using this strategies for response to the BDBEE.

  12. The function of single containment and double containment of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Chen Weijing.

    1985-01-01

    The function and structures of single containment and double containment of PWR nuclear power plant were described briefiy. The dissimilarites of diffent type of containments, which effects the impact of environment are discused. The impact of environment, effected by 'source term', containment gas leak rate and diffusion pattern of the released gas, under different operating condition is analysed. Especially, the impact of environment under LOCA accident is fully analysed

  13. 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

  14. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  15. Dosimetry and fluence calculations on french PWR vessels comparisons between experiments and calculations

    International Nuclear Information System (INIS)

    Nimal, J.C.; Bourdet, L.; Guilleret, J.C.; Hedin, F.

    1988-01-01

    Fluence and damage calculations on PWR pressure vessels and irradiation test specimens are presented for two types of reactor: the franco-belgian (reactor CHOOZ) and the french reactors (CPY program). Comparisons with measurements are given for activation foils and fission detectors; most of them are about irradiation test specimen locations; comparisons are made for the Chooz plant on vessel stainless steel samplings and in the reactor pit

  16. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 4

    International Nuclear Information System (INIS)

    Illy, H.

    1986-09-01

    The 4th supplement of a bibliographical series comprising the analytical and sampling problems of the primary coolant of PWR type reactors covers the literature from 1985 up to July 1986 (220 items). References are listed according to the following topics: boric acid; chloride, chlorine; general; hydrogen isotopes; iodine; iodide; noble gases; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. (V.N.)

  17. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  18. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  19. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  20. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  1. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  2. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  3. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  4. 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

  5. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  6. 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

  7. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  8. 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

  9. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  10. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  11. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  12. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  13. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  14. 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)

  15. First application of hollow fiber filter for PWR condensate polishing

    International Nuclear Information System (INIS)

    Tsuda, S.; Otoha, K.; Takiguchi, H.

    2002-01-01

    In Tsuruga Unit-2 (PWR 1160 MWe commenced commercial operation in 1987), current procedure for secondary system clean-up before start-up had prolonged outage time and had consumed a huge amount of de-ionized (DI) water. In addition, iron oxide in condensate had accelerated the degradation of condensate demineralizer (CD) resin. The corrosion product of iron could also influence the secondary side corrosion of steam generator (SG) tubing if it intruded into SG through CD. To solve these problems, Japan Atomic Power Company (JAPC) decided to introduce hollow fiber filter (HFF) type condensate filter into Tsuruga-2, as the first application to PWR in the world. Because of retro-fitted HFF in Tsuruga Unit-2, limitations for installation space and flow resistance in condensate system and cost reduction required new design for compact and low differential pressure system and for long life filter module. JAPC and ORGANO assessed methodologies to achieve these goals. An advanced HFF system, including a newly developed compact HFF module design, was installed at Tsuruga Unit-2 in 1997 based on the assessment. During the 5 years since the installation, the HFF system has provided excellent crud removal that enables to shorten the outage period and to reduce DI water consumption drastically. Stable differential pressure (dP) trend of the HFF system indicates an expected module life of more than 7 years, with backwash cleaning required only 2 or 3 times per year. In addition to providing the expected operating cost reduction and improved SG tube integrity, numerous additional benefits have resulted from the retrofit. (authors)

  16. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    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 {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  17. Reactor core design calculations and fuel management in PWR; Izracun projekta sredice in upravljanja z forivom tlacnovodnega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    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. Non-intrusive appliance load monitoring system based on a modern kWh-meter

    Energy Technology Data Exchange (ETDEWEB)

    Pihala, H. [VTT Energy, Espoo (Finland). Energy Systems

    1998-12-01

    Non-intrusive appliance load monitoring (NIALM) is a fairly new method to estimate load profiles of individual electric appliances in a small building, like a household, by monitoring the whole load at a single point with one recording device without sub-meters. Appliances have special electrical characteristics, the positive and negative active and reactive power changes during the time they are switched on or off. These changes are called events and are detected with a monitoring device called an event recorder. Different NIALM-concepts developed in Europe and in the United States are generally discussed. The NIALM-concept developed in this study is based on a 3-phase, power quality monitoring kWh-meter and unique load identification algorithms. This modern kWh-meter with a serial data bus to a laptop personal computer is used as die event recorder. The NIALM-concept of this presentation shows for the first time how a kWh-meter can be used at the same time for billing, power quality and appliance end-use monitoring. An essential part of the developed NIALM-system prototype is the software of load identification algorithms which runs in an off-line personal computer. These algorithms are able to identify, with a certain accuracy, both two-state and multi-state appliances. This prototype requires manual-setup in which the naming of appliances is performed. The results of the prototype NIALMS were verified in a large, single family detached house and they were compared to the results of other prototypes in France and the United States, although this comparison is difficult because of different supply systems, appliance stock and number of tested sites. Different applications of NIALM are discussed. Gathering of load research data, verification of DSM-programs, home automation, failure analysis of appliances and security surveillance of buildings are interesting areas of NIALM. Both utilities and customers can benefit from these applications. It is possible to

  19. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  20. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  1. 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)

  2. 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)

  3. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  4. WH-EA: An Evolutionary Algorithm for Wiener-Hammerstein System Identification

    Directory of Open Access Journals (Sweden)

    J. Zambrano

    2018-01-01

    Full Text Available Current methods to identify Wiener-Hammerstein systems using Best Linear Approximation (BLA involve at least two steps. First, BLA is divided into obtaining front and back linear dynamics of the Wiener-Hammerstein model. Second, a refitting procedure of all parameters is carried out to reduce modelling errors. In this paper, a novel approach to identify Wiener-Hammerstein systems in a single step is proposed. This approach is based on a customized evolutionary algorithm (WH-EA able to look for the best BLA split, capturing at the same time the process static nonlinearity with high precision. Furthermore, to correct possible errors in BLA estimation, the locations of poles and zeros are subtly modified within an adequate search space to allow a fine-tuning of the model. The performance of the proposed approach is analysed by using a demonstration example and a nonlinear system identification benchmark.

  5. 700 TWh dry gas for NOK 0,10 per kWh

    International Nuclear Information System (INIS)

    2003-01-01

    In 2002, Norway exported natural gas corresponding to a heat value of 700 TWh. The article looks at what energy price the big gas customers on the continent and in Great Britain really paid for the Norwegian gas. The average price to the wholesale dealer was NOK 0,10/kWh. However, the wholesale dealers quadruple the price out to the end-users. It is also found that if the gas was made available on the home market, then the Norwegian gas distributors would be able to compete on price with the foreign distributors. This fact along with the extreme price on electricity in the winter 2003 gives an impetus to the debate that goes on in Norway on whether or not to build gas power plants

  6. Aqueous phase partitioning of hexachlorocyclohexane (HCH) isomers by biosurfactant produced by Pseudomonas aeruginosa WH-2

    International Nuclear Information System (INIS)

    Sharma, Suman; Singh, Partapbir; Raj, Mayil; Chadha, Bhupinder Singh; Saini, Harvinder Singh

    2009-01-01

    The different isomers of technical-grade hexachlorocyclohexane (t-HCH) including the insecticidal γ-isomer, commonly known as lindane, have been reported to be toxic, carcinogenic and endocrine disrupters. The spatial arrangements of the chlorine atoms on different isomers and low aqueous phase solubility contribute to their persistence in environment, β-HCH being the most resistance to transformation. The biosurfactant preparation of Pseudomonas aeruginosa isolate WH-2 was evaluated for its ability to improve the aqueous phase partitioning of different isomers of HCH-muck. Further, the ability of biosurfactant preparation to emulsify HCH and n-hexadecane was checked under different conditions, usually characteristic of sites contaminated with pollutants viz. wide range of pH, temperature, and salinity. The data obtained from this study will be helpful in designing suitable bioremediation strategies for huge stock piles of HCH-muck and sites polluted by reckless use/disposal of HCH-isomers.

  7. 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

  8. Techniques for Primary-to-Secondary Leak Monitoring in PWR Plants

    International Nuclear Information System (INIS)

    Sohn, Wook; Chi, Jun Hwa; Kang, Duck Won; Tae, Jeong Woo

    2006-01-01

    Historically, corrosion and mechanical damage have made steam generator tubes in PWR plants see various types of degradation from both the primary and secondary sides of the tubes. Since the tube degradation can lead to through-wall failure, the plant personnel should make efforts to prevent the failure. One of such preventive efforts is to monitor primary-to-secondary leakage (PSL) that usually precedes the tube rupture. Thus the objective of PSL monitoring is to make operators to determine when to shutdown the plant in order to minimize the likelihood of propagation of leaks to tube rupture under normal and faulted conditions This paper addresses briefly the status of techniques for PSL monitoring used in PWR plants

  9. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  10. The new lattice code Paragon and its qualification for PWR core applications

    International Nuclear Information System (INIS)

    Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

    2003-01-01

    Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

  11. The empirical intensity of PWR primary coolant pumps failure and repair

    International Nuclear Information System (INIS)

    Milivojevicj, S.; Riznicj, J.

    1988-01-01

    The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

  12. Comparison of the long-time corrosion behavior of certain Zr alloys in PWR, BWR, and laboratory tests

    International Nuclear Information System (INIS)

    Garzarolli, F.; Broy, Y.; Busch, R.A.

    1996-01-01

    Laboratory corrosion tests have always been an important tool for Zr alloy development and optimization. However, it must be known whether a test is representative for the application in-reactor. To shed more light on this question, coupons of several Zr alloys were exposed under isothermal conditions in BWR and PWR type environments. For evaluation of the in-PWR tests and for comparison of out-of-pile and in-pile tests, the different temperatures and times were normalized to a temperature-independent normalized time by assuming an activation temperature (Q/R) of 14,200 K. Comparison of in-PWR and out-of-pile corrosion behavior of Zircaloy shows that corrosion deviates to higher values in PWR if a weight gain of about 50 mg/dm 2 is exceeded. In the case of the Zr2.5Nb alloy, a slight deviation of corrosion as compared to laboratory results starts in PWR only above a weight gain of 100 mg/dm 2 . In BWR, corrosion of Zircaloy is enhanced early in time if compared with out-of-pile. Zr2.5Nb exhibits higher corrosion results in BWR than Zircaloy-4. Alloying chemistry and material condition affect corrosion of Zr alloys. However, several of the material parameters have shown a different ranking in the different environments. Nevertheless, several material parameters influencing in-reactor corrosion like the second phase particle (SPP) size of in-PWR behavior as the Sn and Fe content can be optimized by out-of-pile corrosion tests

  13. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  14. A universal PWR spectral history correction

    International Nuclear Information System (INIS)

    Hutt, P.K.; Nunn, D.L.

    1989-01-01

    The accuracy of a form of universal correction for the difference between depletion conditions assumed in PWR assembly lattice calculations and those experienced in a reactor burn-up is investigated. The correction is based on lattice calculations in which only one such depletion history difference, depletion at two different water densities, is explicitly represented by lattice calculations. The assumption is made that other historical effects bear the same relationship to an appropriate time-average of the two-group neutron flux spectrum. The correction is shown to be accurate for the most important historical effects, depletion with burnable absorbers inserted, control rods inserted or at a different soluble boron level, in addition to density itself. The correction is less accurate for representing depletion at a different fuel or coolant temperature but even in these cases gives an improvement over no correction. In addition it is argued that these historic temperature effects are likely to be of minor importance. (author)

  15. 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

  16. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  17. 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

  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)

    1999-12-31

    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. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

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

  20. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  1. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  2. 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)

  3. 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)

  4. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  5. Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

    International Nuclear Information System (INIS)

    Francis, A.J.

    1990-01-01

    It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

  6. 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.

  7. Qualification tests for PWR control element drive mechanism

    International Nuclear Information System (INIS)

    Kim, In Yong; Jin, Choon Eon; Choi Suhn

    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. Changes in 900 MW PWR alarm processing policy

    Energy Technology Data Exchange (ETDEWEB)

    Pont, M [Electricite de France, Generation and Transmission, Nuclear Power Plant Operations, Paris (France)

    1997-09-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs.

  9. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Pont, M.

    1997-01-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  10. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  11. Dose rate evaluation after accident in a PWR

    International Nuclear Information System (INIS)

    Cladel, C.; Duchemin, B.; Le Dieu de Ville, A.; Nimal, B.; Nimal, J.C.; Evrard, J.M.

    1983-05-01

    A calculation scheme for the gamma radiation dose rate after accident in a PWR is presented. These studies use a fine description of the geometry and of the fission product inventory. Some results are given and some improvements are planned

  12. 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.

  13. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  14. Per maand of per kWh?: beoordeling van de effecten van een ander tariefsysteem voor kleinverbruikers electriciteit

    NARCIS (Netherlands)

    Aalbers, R.F.A.; de Nooij, M.

    2006-01-01

    Dit rapport bepaalt de effecten van de invoering van een capaciteitstarief voor de kosten van het transport van elektriciteit voor kleinverbruikers. Centraal kenmerk van een capaciteitstarief is dat kleinverbruikers niet langer per afgenomen kWh, maar een vast bedrag per aansluiting betalen. In het

  15. Effects of Wh-Question Graphic Organizers on Reading Comprehension Skills of Students with Autism Spectrum Disorders

    Science.gov (United States)

    Bethune, Keri S.; Wood,Charles L.

    2013-01-01

    Students with autism spectrum disorders often have difficulty with reading comprehension. This study used a delayed multiple baseline across participants design to evaluate the effects of graphic organizers on the accuracy of wh-questions answered following short passage reading. Participants were three elementary-age students with autism spectrum…

  16. 29 CFR Appendix G to Part 825 - Certification of Qualifying Exigency for Military Family Leave (Form WH-384)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 3 2010-07-01 2010-07-01 false Certification of Qualifying Exigency for Military Family Leave (Form WH-384) G Appendix G to Part 825 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR OTHER LAWS THE FAMILY AND MEDICAL LEAVE ACT OF 1993 Pt. 825, App. G...

  17. 29 CFR Appendix E to Part 825 - Designation Notice to Employee of FMLA Leave (Form WH-382)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 3 2010-07-01 2010-07-01 false Designation Notice to Employee of FMLA Leave (Form WH-382) E Appendix E to Part 825 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR OTHER LAWS THE FAMILY AND MEDICAL LEAVE ACT OF 1993 Pt. 825, App. E Appendix E to Part...

  18. 29 CFR Appendix C to Part 825 - Notice to Employees Of Rights Under FMLA (WH Publication 1420)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 3 2010-07-01 2010-07-01 false Notice to Employees Of Rights Under FMLA (WH Publication 1420) C Appendix C to Part 825 Labor Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR OTHER LAWS THE FAMILY AND MEDICAL LEAVE ACT OF 1993 Pt. 825, App. C Appendix C to Part...

  19. 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

  20. 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

  1. Application of directional solidification ingot (LSD) in forging of PWR reactor vessel heads

    International Nuclear Information System (INIS)

    Benhamou, C.; Poitrault, I.

    1985-09-01

    Creusot-Loire Industrie uses this type of ingot for manufacture of Framatome 1300 and 1450 MW 4-loop PWR reactor vessel heads. This type of ingot offers a number advantages: improved internal soundness; greater chemical, structural and mechanical homogeneity of the finished part; simplified forging process. After a brief description of the pouring and solidification processes, this paper presents an analysis of the results of examinations performed on the prototype forging, as well as review of results obtained during industrial fabrication of dished heads from LSD ingots. The advantages of the LSD ingot over conventional ingots are discussed in conclusion

  2. Seismic analysis of the reactor coolant system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Borsoi, L.; Sollogoub, P.

    1986-01-01

    For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

  3. Search for the SM Higgs Boson in the Channel $WH \\to l\

    CERN Document Server

    Will, Jonas Zacharias

    One of the most important scientific challenges of ATLAS and CMS, multi-purpose de- tectors at CERN’s Large Hadron Collider (LHC), is the discovery or exclusion of the longly sought standard model Higgs boson predicted almost fifty years ago. In summer 2012, both ATLAS and CMS discovered a new particle. Its mass is determined to be 126 . 0 ± 0 . 4 (stat) ± 0 . 4 (sys) GeV (ATLAS) and 125 . 3 ± 0 . 4 (stat) ± 0 . 5 (sys) GeV (CMS) [ 1 , 2 ]. Its further properties are so far consistent with the predicted properties of a standard model Higgs boson within large uncertainties. Besides the Higgs search in the sensitive bosonic channels, H → γγ , H → ZZ , and H → WW , the fermionic channels H → ττ and H → b b contributed to the exclusion of a standard model Higgs boson below the observed excess and are essential for measuring the couplings of the new particle to fermions. In the analysis presented here, the associated Higgs production WH in the Higgs decay channel H → b b is studied on the co...

  4. Suncor to buy 350,000 kWh/y of wind

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

    An agreement between Suncor Energy Inc., and Vision Quest Windelectric Inc., of Calgary was announced whereby Suncor will fund the generation of up to 350,000 kWh per year of electricity from wind turbines in Alberta. As a result of this agreement, 350,000 kgs of greenhouse gases will be avoided, a modest, but not unimportant reduction in air emissions. Vision Quest recently developed two new wind energy facilities near Pincher Creek and Hillspring, Alberta, to serve the new market for emission-free electricity. The addition of new wind turbine facilities is planned to serve Suncor and other participants' requirement. Vision Quest's wind turbine power plants have been licensed under Environment Canada's Environmental Choice Program, verifying that the power is produced from emission-free sources. In another development, Suncor and Calgary-based Mercury Electric Corporation recently announced a pilot project that will see Suncor's conventional oil and gas business use mini-turbines to generate electricity from flare gas. Mercury will test the mini-turbine technology at a Suncor facility at Joffre near Red Deer, Alberta. Beyond these modest efforts, Suncor is committed to action on the global climate issue by continuing to manage the company's own greenhouse gas emissions and their impact. Suncor is developing renewable sources of energy, supporting environmental and economic research, pursuing domestic and international offsets, providing constructive public policy input, educating and engaging stakeholders and reporting on the company's progress in the environmental protection area

  5. 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)

  6. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-01-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  7. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Moreira, F.J.

    1984-01-01

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

  8. Chemical process measurements in PWR-type nuclear power plants

    International Nuclear Information System (INIS)

    Glaeser, E.

    1978-01-01

    In order to achieve high levels of availability of nuclear power plants equipped with pressurized water reactors, strict standards have to be applied to the purity of coolant and of other media. Chemical process measurements can meet these requirements only if programmes are established giving maximum information with minimum expenditure and if these programmes are realized with effective analytical methods. Analysis programmes known from literature are proved for their usefulness, and hints are given for establishing rational programmes. Analytical techniques are compared with each other taking into consideration both methods which have already been introduced into nuclear power plant practice and methods not yet generally used in practice, such as atomic absorption spectrophotometry, gas chromatography, etc. Finally, based on the state of the art of chemical process measurements in nuclear power plants, the trends of future development are pointed out. (author)

  9. 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

  10. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  11. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  12. Performance of PWR study in the technology supplier countries: south korea and japan case

    International Nuclear Information System (INIS)

    Sriyana

    2007-01-01

    Electricity is needed as an infrastructure to support the national economic growth. For economic development sustainability, energy alternatives should be provided. Nuclear Power Plant (NPP) become the alternative of electricity generation for optimum energy mix in Indonesia and planned to operate in the 2016. Several studies have already done to prepare the NPP construction. This study focused on NPP performance especially PWR type in Asia, namely Japan and South Korea. Methodology used in this is literature tracing and a small calculation. The energy availability per unit per year is used as a parameter for evaluating the NPP performance. This conclusion are 1) the amount of NPP - PWR type in Japan is 22 units with total operational experiences 526 reactor-years and the average energy availability factor about 70.7% per unit per year. Meanwhile for the same type South Korea has 16 unit with total operational experience 222 reactor-years and average availability factor per unit per year is about 86.9%. 2) the 1000 class of PWR type both South Korea and Japan have 14 units. The operational experiences for thi class is 170 reactor-year for South Korean and 307 reactor-year for Japan. Meanwhile the average availability factor per unit per year is about 87.0% for South Korea and 69.6% for Japan. 3) the average availability factor is closed to capacity factor, so is important for real figure in assuming the techno-economic parameters, because it will influence the result o economic calculation. (author)

  13. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  14. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  15. Development of forging technology for PWR primary piping

    International Nuclear Information System (INIS)

    Morin, F.; Badeau, J.P.; Lambs, R.

    1996-01-01

    The purpose of this presentation is to give information on the changes in the design and manufacture of Primary Piping for electronuclear boilers of the Pressurized Water Reactor type (PWR) which has resulted in the making of one-piece forged lines including stub pipes and arcs. The optimization of these items is aimed at improving the life of the new power stations as well as guaranteeing their safety, while reducing inspection and maintenance requirements in service. The demonstration of the manufacturing feasibility has just been completed. It has taken material form in the installation, on the CIVAUX 1 section, of the first one-piece cold leg in the world. It will shortly be followed by the installation on the CIVAUX 2 section of a complete loop of bent forged pipes. Therefore, this new know-how is going to be incorporated in the French Rules (RCC-M) and can be directly taken into consideration both in the next work to be done and in the design and definition of a future nuclear reactor

  16. Improving MODPRESS heat loss calculations for PWR pressurizers

    International Nuclear Information System (INIS)

    Ramos, Natalia V.; Lira, Carlos A. Brayner O.; Castrillho, Lazara S.

    2009-01-01

    The improvement of heat loss calculations in MODPRESS transient code for PWR pressurizer analysis is the main focus of this investigation. Initially, a heat loss model was built based on heat transfer coefficient (HTC) correlations obtained in handbooks of thermal engineering. A hand calculation for Neptunus experimental test number U47 yielded a thermal power loss of 11.2 kW against 17.3 kW given by MODPRESS at the same conditions, while the experimental estimate is given as 17 kW. This comparison is valid only for steady state or before starting the transient experiment, because MODPRESS does not update HTC's when the transient phase begins. Furthermore, it must be noted that MODPRESS heat transfer coefficients are adjusted to reproduce the experimental value of the specific type of pressurizer. After inserting the new routine for HTC's into MODPRESS, the heat loss was calculated as 11.4 kW, a value very close to the first estimate but far below 17 kW found in the U47 experiment. In this paper, the heat loss model and results will be described. Further research is being developed to find a more general HTC that allows the analysis of the effects of heat losses on transient behavior of Neptunus and IRIS pressurizers. (author)

  17. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Ball, D.G.

    1983-01-01

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  18. 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)

  19. "Which mouse kissed the frog?" Effects of age of onset, length of exposure, and knowledge of case marking on the comprehension of wh-questions in German-speaking simultaneous and early sequential bilingual children.

    Science.gov (United States)

    Roesch, Anne Dorothee; Chondrogianni, Vasiliki

    2016-05-01

    Studies examining age of onset (AoO) effects in childhood bilingualism have provided mixed results as to whether early sequential bilingual children (eL2) differ from simultaneous bilingual children (2L1) and L2 children on the acquisition of morphosyntax. Differences between the three groups have been attributed to other factors such as length of exposure (LoE), language abilities, and the phenomenon to be acquired. The present study investigates whether four- to five-year-old German-speaking eL2 children differ from 2L1 children on the acquisition of wh-questions, and whether these differences can be explained by AoO, LoE, and/or knowledge of case marking. The 2L1 children outperformed the eL2 children in terms of accuracy; however, both bilingual groups exhibited similar error patterns. This suggests that 2L1 and eL2 bilingual children are sensitive to the same morphosyntactic cues, when comprehending wh-questions. Finally, children's performance on the different types of wh-questions was explained by a combination of knowledge of case marking, LoE, and AoO.

  20. The effects of L2 proficiency level on the processing of wh-questions among Dutch second language speakers of English

    NARCIS (Netherlands)

    Jackson, C.N.; Hell, J.G. van

    2011-01-01

    Using a self-paced reading task, the present study explores how Dutch-English L2 speakers parse English wh-subject-extractions and wh-object-extractions. Results suggest that English native speakers and highly-proficient Dutch–English L2 speakers do not always exhibit measurable signs of on-line

  1. 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

  2. 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.

  3. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  4. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  5. 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.

  6. Summary of PWR leak detection studies

    International Nuclear Information System (INIS)

    Cho, J.H.; Elia, F.A. Jr.

    1986-01-01

    Thermal-hydraulic analysis can be used to determine the location and magnitude of leaks inside and location of leaks outside a pressurized water reactor (PWR) containment as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside containment. Plant-specific analyses are utilized to predict change in parameters such as local dew point temperature, relative humidity, dry bulb temperature, and flow rate to sump for various leak rates and enthalpies. These parameter responses are then programmed into the plant computer and instrumentation is provided for area monitoring. The actual inputs are continuously monitored and compared to the predicted plant responses to identify the leak location and quantify the leak. This study concludes that a system that monitors dew point (or relative humidity) and dry bulb temperature changes together with the flow rate to the sump will provide the capability to both locate and quantify a leak inside a containment, while a system that monitors dew point temperature (or relative humidity) changes will provide the capability to locate a leak outside a containment

  7. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  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. 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

  10. 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.

  11. Multicriteria analysis of public protection in PWR's

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    In order to manage a risk efficiently and to reach the ALARA level of protection, the best possible protection options must be employed. As the available resources are limited, it is not always possible to choose those options that minimize the risk, therefore a compromise must be made between risks and safety expenses. When the choice is difficult or complex, finding such a compromise can be facilitated by resorting to a decision aiding method which allows the assessment of the respective advantages of the various protection options considered. The multicriteria methods employ successive comparisons. Instead of searching for a final indicator expressing the performance of each option they compare all option pairs in order to determine if the gap between their respective advantages and disadvantages is sufficient to estimate that one option of the each pair is better than the other. Instead of judging each option globally these methods evaluate the advantages and disadvantages associated with the eventual choice of an option as compared with the others. These differential and comparative approach gives more flexibility and allows the introduction of qualitative criteria. The method presented here (Electre 3), one of the most recent ones, allows a multicriteria comparison of a set of options keeping into account the uncertainties associated with the options or the preferences. In order to illustrate this method a simple example (4 options, 4 criteria) dealing with a PWR liquid releases treatment system, is taken up

  12. Modeling of PWR fuel at extended burnup

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  13. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  14. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael M.; Silva, Antonio Teixeira

    2015-01-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  15. 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)

  16. Knowledge of ageing phenomenons of materials used in the PWR power plants

    International Nuclear Information System (INIS)

    Vancon, D.; Meyzaud, Y.; Soulat, P.

    1996-01-01

    The nuclear power plants with PWR type reactors are planned to work during forty years and are the subject of studies aiming to check their integrity during all their life. The materials used to the fabrication of the components can be submitted different stress. The temperature, the mechanical constraints, the irradiation are examples of stress which can make the materials getting old. This text presents three themes: the ageing by irradiation, the thermal ageing and the corrosion, and their principle industrial consequences. (N.C.)

  17. Effect of the Th-232/U-238 ratio on the conversion ratio of PWR's

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    The conversion ratio, the consumption of natural uranium and separative work for current PWR lattice designs are studied as function of: the initial fraction of thorium in the fuel, the type of fuel recycling employed, and the growth rate of the system of reactors. The incentive to utilize thorium is maximum when: thorium is used together with fully enriched uranium (93%); if only uranium can be recycled; and the system of reactors is at equilibrium. Under these conditions, the comsumption of natural uranium and separative work are lower than the all-uranium-fuel case by 41% and 18%,respectively [pt

  18. Development of emergency operator support system for next Japanese PWR plants

    International Nuclear Information System (INIS)

    Ito, K.; Hanada, S.; Yoshida, Y.; Sugino, K.

    2006-01-01

    The purpose of main control room improvement is to reduce operator workload and potential human errors by offering a better working environment where operators can maximize their abilities. Japanese PWR utilities and Mitsubishi have developed an operator support system entitled Emergency Operator Support System (EOSS). The system supports operators in incidental/accidental situations which may be worsened by human errors. In order to confirm the validity of the system, a proto type was built, and was evaluated by operator crews. The consequence showed good result of effectiveness in avoiding potential human errors and decreasing workload of operators. (authors)

  19. Application of tearing modulus stability concepts to nuclear piping. Final report. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cotter, K.H.; Chang, H.Y.; Zahoor, A.

    1982-02-01

    The recently developed tearing modulus stability concept was successfully applied to several boiling water reactor (BWR) and pressurized water reactor (PWR) piping systems. Circumferentially oriented through-the-thickness cracks were postulated at numerous locations in each system. For each location, the simplified tearing stability methods developed in USNRC Report NUREG/CR-0838 were used to determine crack stability. The J-T diagram was used to present the results of the computations. The piping systems considered included Type 304 stainless steel as well as A106 carbon steel materials. These systems were analyzed using the piping analysis computer code MINK.

  20. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    International Nuclear Information System (INIS)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-01-01

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation

  1. Development of automated generation system of accidental operating procedures for a PWR

    International Nuclear Information System (INIS)

    Artaud, J.L.

    1991-06-01

    The aim of the ACACIA project is to develop an automated generation system of accident operating procedures for a PWR. This research and development study, common at CEA and EDF, has two objectives: at mean-dated the realization of a validation tool and a procedure generation; at long-dated the dynamic generation of real time procedures. This work is consecrated at the realization of 2 prototypes. These prototypes and the technics used are described in detail. The last chapter explores the perspectives given by this type of tool [fr

  2. 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

  3. 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

  4. Oligo-cyclic damage and behaviour of a 304 L austenitic stainless steel according to environment (vacuum, air, PWR primary water) at 300 C

    International Nuclear Information System (INIS)

    De Baglion, L.

    2011-01-01

    Nowadays, for nuclear power plants licensing or operating life extensions, various safety authorities require the consideration of the primary water environment effect on the fatigue life of Pressurized Water Reactor (PWR) components. Thus, this work focused on the study of low cycle fatigue damage kinetics and mechanisms, of a type 304L austenitic stainless steel. Several parameters effects such as temperature, strain rate or strain amplitude were investigated in air as in PWR water. Thanks to targeted in-vacuum tests, the intrinsic influence of these parameters and environments on the fatigue behaviour of the material was studied. It appears that compared with vacuum, air is already an active environment which is responsible for a strong decrease in fatigue lifetime of this steel, especially at 300 C and low strain amplitude. The PWR water coolant environment is more active than air and leads to increased damage kinetics, without any modifications of the initiation sites or propagation modes. Moreover, the decreased fatigue life in PWR water is essentially attributed to an enhancement of both initiation and micropropagation of 'short cracks'. Finally, the deleterious influence of low strain rates on the 304L austenitic stainless steel fatigue lifetime was observed in PWR water environment, in air and also in vacuum without any environmental effects. This intrinsic strain rate effect is attributed to the occurrence of the Dynamic Strain Aging phenomenon which is responsible for a change in deformation modes and for an enhancement of cracks initiation. (author)

  5. Complete genome sequence of Agarivorans gilvus WH0801(T), an agarase-producing bacterium isolated from seaweed.

    Science.gov (United States)

    Zhang, Pujuan; Rui, Junpeng; Du, Zongjun; Xue, Changhu; Li, Xiangzhen; Mao, Xiangzhao

    2016-02-10

    Agarivorans gilvus WH0801(T), an agarase-producing bacterium, was isolated from the surface of seaweed. Here, we present the complete genome sequence, which consists of one circular chromosome of 4,416,600 bp with a GC content of 45.9%. This genetic information will provide insight into biotechnological applications of producing agar for food and industry. Copyright © 2015 Elsevier B.V. All rights reserved.

  6. Liquid radioactive waste processing improvement of PWR nuclear power plants; Melhorias no processamento de rejeitos liquidos radioativos de usinas nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: wolter@eletronuclear.gov.br; monteiro@peq.coppe.ufrj.br; aquilinosenra@lmp.ufrj.br

    2005-07-01

    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)

  7. Proteomic responses of oceanic Synechococcus WH8102 to phosphate and zinc scarcity and cadmium additions

    Directory of Open Access Journals (Sweden)

    Alysia eCox

    2013-12-01

    Full Text Available Synechococcus sp. WH 8102 is a motile marine cyanobacterium isolated originally from the Sargasso Sea. To test the response of this organism to cadmium (Cd -generally considered a toxin- cultures were grown in a matrix of high and low zinc (Zn and phosphate (PO43- and were then exposed to an addition of 4.4 pM free Cd2+ at mid-log phase and harvested after 24 h. Whereas Zn and PO43- had little effect on overall growth rates, in the final 24 h of the experiment three growth effects were noticed: i low PO43- treatments showed increased growth rates relative to high PO43- treatments, ii the Zn/high PO43- treatment appeared to enter stationary phase, and iii Cd increased growth rates further in both the low PO43- and Zn treatments. Global proteomic analysis revealed that: i Zn appeared to be critical to the PO43- response in this organism, ii bacterial metallothionein (SmtA appears correlated with PO43- stress-associated proteins, iii Cd has the greatest influence on the proteome at low PO43- and Zn, iv Zn buffered the effects of Cd, and v in the presence of both replete PO43- and added Cd the proteome showed little response to the presence of Zn. Similar trends in alkaline phosphate (ALP and SmtA suggest the possibility of a Zn supply system to provide Zn to ALP that involves SmtA. In addition, proteome results were consistent with a previous transcriptome study of PO43- stress (with replete Zn in this organism, including the greater relative abundance of ALP (PhoA, ABC phosphate binding protein (PstS and other proteins. Yet with no Zn in this proteome experiment the PO43- response was quite different including the greater relative abundance of five hypothetical proteins with no increase in PhoA or PstS, suggesting that Zn nutritional levels are connected to the PO43- response in this cyanobacterium. Alternate ALP PhoX (Ca was found to be a low abundance protein, suggesting that PhoA (Zn, Mg may be more environmentally relevant than PhoX.

  8. 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

  9. Crack growth testing of cold worked stainless steel in a simulated PWR primary water environment to assess susceptibility to stress corrosion cracking

    International Nuclear Information System (INIS)

    Tice, D.R.; Stairmand, J.W.; Fairbrother, H.J.; Stock, A.

    2007-01-01

    Although austenitic stainless steels do not show a high degree of susceptibility to stress corrosion cracking (SCC) in PWR primary environments, there is limited evidence from laboratory testing that crack propagation may occur under some conditions for materials in a cold-worked condition. A test program is therefore underway to examine the factors influencing SCC propagation in good quality PWR primary coolant. Type 304 stainless steel was subjected to cold working by either rolling (at ambient or elevated temperature) or fatigue cycling, to produce a range of yield strengths. Compact tension specimens were fabricated from these materials and tested in simulated high temperature (250-300 o C) PWR primary coolant. It was observed that the degree of crack propagation was influenced by the degree of cold work, the crack growth orientation relative to the rolling direction and the method of working. (author)

  10. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    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).

  11. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  12. PWR and BWR light water reactor systems in the USA and their fuel cycle

    International Nuclear Information System (INIS)

    Crawford, W.D.

    1977-01-01

    Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

  13. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  14. 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

  15. 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)

  16. 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

  17. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  18. 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)

  19. 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)

  20. A literature review: addressing indigenous parental substance use and child welfare in Aotearoa: a Whānau Ora framework.

    Science.gov (United States)

    McLachlan, Andre; Levy, Michelle; McClintock, Kahu; Tauroa, Roimata

    2015-01-01

    Parental substance use disorders (SUDs) for Māori, the indigenous people of Aotearoa/New Zealand and an ethnic minority, are considered to be contributors to adverse effects on outcomes for their children. This article offers a review of international and Aotearoa literature in regard to key considerations for Māori parents with SUDs who present to an Alcohol and Drug specialist for assessment and treatment. Factors to increase positive outcomes for Māori children of parents with SUDs are promoted. Effective adult AoD services provide support to parents with SUDs through comprehensive assessment and intervention plans that consider both individual and familial risk and protective factors. In this context, it is imperative that possible child welfare issues are identified early to ensure prevention or intervention. The AoD workforce must have the knowledge and skills to facilitate access to other relevant sectors, such as education, employment, and housing. An AoD workforce that is effective with Māori must not only have these abilities, but also have at least some basic knowledge and skills in Whānau Ora philosophy and Whānau-centered best practice. To address these processes, AoD specialist services need to acquire a set of knowledge and skills. These include increasing the knowledge and skills associated with the realities of lifestyles centered in low socioeconomic communities and co-occurring issues that contribute to poor health outcomes. To assist Māori, several key processes are proposed. This includes working in a Whānau-centered approach with Whānau as a collective entity, based on Māori foundations; understanding intergenerational dynamics; and endorsing a group capacity for self-determination. Research and training in Whanau ora philosophy and Whānau-centered best practices will be essential for developing an appropriate AoD workforce, which would provide the foundations for improving AoD service delivery for Māori parents with SUDs.

  1. 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

  2. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  3. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  4. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  5. 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

  6. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  7. 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)

  8. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

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

  9. 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

  10. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  11. 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.

  12. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  13. 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

  14. Enlarged level-1 PSA in regard to assessment of cross-cutting effects of hazards and consideration of their uncertainties for a KONVOI type PWR reference plant; Erweiterte PSA der Stufe 1 im Hinblick auf die Behandlung uebergreifender Einwirkungen und die Beruecksichtigung ihrer Unsicherheiten am Beispiel einer Anlage vom Typ Konvoi

    Energy Technology Data Exchange (ETDEWEB)

    Haider, C.; Hofer, E.; Kloos, M.; Kuntze, W.; Liemersdorf, H.; Roewekamp, M.; Schwinges, B.; Tuerschmann, M.; Brenig, H.W.; Sommerfeld, H.

    2001-06-01

    In the frame of supporting BMU regarding generic questions concerning probabilistic safety analyses for nuclear power plants as well as regarding evaluation of nuclear specific standards and guidelines the significant contributions to damage states resulting from plant internal and external hazards had to be estimated for a German Konvoi type PWR reference plant. Furthermore, the suitability of the available methods for assessing these hazards should be checked. In the report presented hereafter, only the plant internal hazard Fire out of all the hazards to be considered was probabilistically analysed in detail First of all, screening analyses were carried out for identifying relevant plant areas and to assess their respective efficiency for a proper selection procedure. For a selected, plant area identified to be relevant (area of the cable distributions inside the reactor containment) an indepth analysis was performed. This analysis included all the steps of the probabilistic assessment, starting from the estimation of the fire occurrence frequency, followed by investigations on the fire effects and fire propagation, up to the determination of the fire induced failure probabilities of safety related equipment including the consequences on systems. In addition, the analyses contained particular uncertainty and sensitivity studies, for which aleatoric and epistemic uncertainties were distinguished. As a result of the screening analyses as well as of the in-depth investigations regarding the fire hazard, no significant contributions for the total frequencies for system, core, or plant damage states have been found. In this context, it has to be noticed that the study presented hereafter does not cover a complete fire PSA. With respect to assessing the available methods it has been found that improvements concerning the screening process as well as concerning the probabilistic fire event tree analyses are necessary. With regard to further hazards, a site specific

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  17. Alternative water chemistry for the primary loop of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Henzel, N [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Advanced fuel element concepts (longer cycles, higher burnup, increased rod power) call for more reactivity binding capacity and, moreover, might produce higher void fractions, particularly in the hot channel. Thus, on the one hand, more alcalizing agent is needed to maintain a high coolant pH according to the approved ``modified boron-lithium mode of operation`` in the presence of more boric acid (chemical shim); on the other hand, increasing enrichment of coolant constituents due to local boiling (higher void fraction), which must not result in accelerated corrosion of fuel cladding and structural materials, imposes enhanced requirements on both, materials technology and water chemistry. At present, the use of boric acid enriched in B10 (the isotope effective in terms of reactivity control) appears to advantageously compromise in capturing more neutrons with less total boron while maintaining or even slightly reducing lithium concentrations at the same time. There is no feasible alternative for boric acid used as the chemical shim and for hydrogen gas as the reducing agent used to suppress oxygen formation by water radiolysis. Systematic screening as performed in phase 1 of a recent project proved potassium hydroxide to be the only potential candidate to favourably replace lithium 7 hydroxide as an alcalizing agent. Unfortunately, the results of pertinent comparative corrosion tests are not unambiguous, and available operational experience with potassium hydroxide in WWER plants is not readily applicable to western world-type PWR plants. Therefore, a switch-over from lithium to potassium can be envisaged only subsequent to a comprehensive qualification program which is planned to be the objective of phase 2 of the project. This program should also comprise zinc addition tests in order to confirm the alleged positive impact of this element on corrosion rates and activity buildup. (Abstract Truncated)

  18. 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

  19. 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

  20. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  1. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  2. 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

  3. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    International Nuclear Information System (INIS)

    Saernmark, Ivan; Lenz, Herbert

    2008-01-01

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction

  4. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    Energy Technology Data Exchange (ETDEWEB)

    Saernmark, Ivan; Lenz, Herbert [WesDyne TRC AB, Stockholm (Sweden)

    2008-04-15

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction.

  5. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  6. Effect of hardening on the crack growth rate of austenitic stainless steels in primary PWR conditions

    International Nuclear Information System (INIS)

    Castano, M.L.; Garcia, M.S.; Diego, G. de; Gomez-Briceno, D.; Francia, L.

    2002-01-01

    Intergranular cracking of non-sensitized materials, found in light water reactor (LWR) components exposed to neutron radiation, has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). Cracking of baffle former bolts, fabricated of AISI-316L and AISI-347, have been reported in some Europeans and US PWR plants. Examinations of removed bolts indicate the intergranular cracking characteristics can be associated with IASCC phenomena. Neutron radiation produce critical modifications of the microstructure and microchemical of stainless steels such hardening due to irradiation and Radiation Induce Segregation (RIS) at grain boundaries, among others. Chromium depletion at grain boundary due to RIS seems to justify the intergranular cracking of irradiated materials, both in plant and in lab tests, at high electrochemical corrosion potential (BWR-NWC environments), but it is not enough to explain cracking at low corrosion potential (BWR-HWC and PWR environments). In these latter conditions, hardening is considered a possible additional mechanism to explain the behavior of irradiated material. Radiation Hardening can be simulated in non irradiated material by mechanical deformation. Although some differences exists in the types of defects produced by radiation and mechanical deformation, it is accepted that the study of the stress corrosion behavior of unirradiated austenitic steels with different hardening levels would contribute to the understanding of IASCC mechanism. In order to evaluate the influence of hardening on the stress corrosion susceptibility of austenitic steels, crack growth rate tests with 316L and 347 stainless steels with nominal yield strengths from 500 to 900 MPa, produced by cold work are being carried out at 340 deg C in PWR conditions. Preliminary results indicate that crack propagation was obtained in the 316Lss and 347ss cold worked, even with a yield strength of 550 MPa. (authors)

  7. Physics of plutonium and americium recycling in PWR using advanced fuel concepts

    International Nuclear Information System (INIS)

    Hourcade, E.

    2004-01-01

    PWR waste inventory management is considered in many countries including Frances as one of the main current issues. Pu and Am are the 2 main contents both in term of volume and long term radio-toxicity. Waiting for the Generation IV systems implementation (2035-2050), one of the mid-term solutions for their transmutation involves the use of advanced fuels in Pressurized Water Reactors (PWR). These have to require as little modification as possible of the core internals, the cooling system and fuel cycle facilities (fabrication and reprocessing). The first part of this paper deals with some neutronic characteristics of Pu and/or Am recycling. In a second part, 2 technical solutions MOX-HMR and APA-DUPLEX-84 are presented and the third part is devoted to the study of a few global strategies. The main neutronic parameters to be considered for Pu and Am recycling in PWR are void coefficient, Doppler coefficient, fraction of delayed neutrons and power distribution (especially for heterogeneous configurations). The modification of the moderation ratio, the opportunity to use inert matrices (targets), the optimisation of Uranium, Plutonium and Americium contents are the key parameters to play with. One of the solutions (APA-DUPLEX-84) presented here is a heterogeneous assembly with regular moderation ratio composed with both target fuel rods (Pu and Am embedded in an inert matrix) and standard UO 2 fuel rods. An EPR (European Pressurised Reactor) type reactor, loaded only with assemblies containing 84 peripheral targets, can reach an Americium consumption rate of (4.4; 23 kg/TWh) depending on the assembly concept. For Pu and Am inventories stabilisation, the theoretical fraction of reactors loaded with Pu + Am or Pu assemblies is about 60%. For Americium inventory stabilisation, the fraction decreases down to 16%, but Pu is produced at a rate of 18.5 Kg/TWh (-25% compared to one through UOX cycle)

  8. Fatigue life evaluation method of austenitic stainless steel in PWR water

    International Nuclear Information System (INIS)

    Sakaguchi, Katsumi; Nomura, Yuichiro; Suzuki, Shigeki; Kanasaki, Hiroshi; Higuchi, Makoto

    2006-09-01

    It is known that the fatigue life in elevated temperature water is substantially reduced compared with that in the air. The fatigue life reduction has been investigated experimentally in EFT project of Japan Nuclear Energy Safety Organization (JNES) to evaluate the environmental effect on fatigue life. Many tests have been done for carbon, low alloy, stainless steels and nickel-based alloy under the various conditions. In this paper, the results of the stainless steel in simulated PWR water environments were reported. Fatigue life tests in simulated PWR environments were carried out and the effect of key parameters on fatigue life reduction was examined. The materials used in this study were base and weld metal of austenitic stainless steel SS316, weld metal of SS304 and the base and aged metal of the duplex stainless steel SCS14A. In order to evaluate the effects of stain amplitude, strain rate, strain ratio, temperature, aging, water flow rate and strain holding time, many fatigue tests were examined. In transient condition in an actual plant, however, such parameters as temperature and strain rate are not constant. In order to evaluate fatigue damage in actual plant on the basis of experimental results under constant temperature and strain rate condition, the modified rate approach method was developed. Various kinds of transient have to be taken into account of in actual plant fatigue evaluation, and stress cycle of several ranges of amplitude has to be considered in assessing damage from fatigue. Generally, cumulative usage factor is applied in this type of evaluation. In this study, in order to confirm the applicability of modified rate approach method together with cumulative usage factor, fatigue tests were carried out by combining stress cycle blocks of different strain amplitude levels, in which strain rate changes in response to temperature in a simulated PWR water environment. Consequently, fatigue life could be evaluated with an accuracy of factor of 3

  9. Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

    International Nuclear Information System (INIS)

    Cho, Chun-Hyung; Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun

    2011-01-01

    Research highlights: → We compare the costs of wet and dry interim storage facilities for PWR spent fuel. → We use the parametric method and quotations to deduce unknown cost items. → Net present values and levelized unit prices are calculated for cost comparisons. → A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

  10. Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chun-Hyung, E-mail: skycho@krmc.or.kr [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2011-05-15

    Research highlights: > We compare the costs of wet and dry interim storage facilities for PWR spent fuel. > We use the parametric method and quotations to deduce unknown cost items. > Net present values and levelized unit prices are calculated for cost comparisons. > A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

  11. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

    International Nuclear Information System (INIS)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

    2013-10-01

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

  12. A Hold-down Margin Assessment using Statistical Method for the PWR Fuel Assembly

    International Nuclear Information System (INIS)

    Jeon, S. Y.; Park, N. K.; Lee, K. S.; Kim, H. K.

    2007-01-01

    The hold-down springs provide an acceptable hold down force against hydraulic uplift force absorbing the length change of the fuel assembly relative to the space between the upper and lower core plates in PWR. These length changes are mainly due to the thermal expansion, irradiation growth and creep down of the fuel assemblies. There are two kinds of hold-down springs depending on the different design concept of the reactor internals of the PWR in Korea, one is a leaf-type hold down spring for Westinghouse type plants and the other is a coil-type hold-down spring for OPR1000 (Optimized Power Reactor 1000). There are four sets of hold-down springs in each fuel assembly for leaf type hold-down spring and each set of the hold-down springs consists of multiple tapered leaves to form a cantilever leaf spring set. The length, width and thickness of the spring leaves are selected to provide the desired spring constant, deflection range, and hold down force. There are four coil springs in each fuel assembly for coil-type hold-down spring. In this study, the hold-down forces and margins were calculated for the leaf-type and coil-type hold-down springs considering geometrical data of the fuel assembly and its components, length changes of the fuel assembly due to thermal expansion, irradiation growth, creep, and irradiation relaxation. The hold-down spring forces were calculated deterministically and statistically to investigate the benefit of the statistical calculation method in view of hold-down margin. The Monte-Carlo simulation method was used for the statistical hold down force calculation

  13. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

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

  14. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

  15. Method of characteristics - Based sensitivity calculations for international PWR benchmark

    International Nuclear Information System (INIS)

    Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

    2013-01-01

    Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

  16. 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

  17. 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

  18. 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

  19. Is it possible to improve regulation system of PWR

    International Nuclear Information System (INIS)

    Bonnemay, A.; Martinez, J.M.

    1983-03-01

    This paper deals with two problems: first of all, it presents the critical analysis of usually implemented general regulation systems, on PWR plants, and derives from it same possibilities to improve the transient behavior of reactor, the second part is a proposition from an automatic control system for spatial distribution of flux

  20. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)