WorldWideScience

Sample records for nuclear containment building

  1. Vulnerability analysis in a pwr nuclear power plant containment building

    OpenAIRE

    Musolas Otaño, Antoni Maria

    2013-01-01

    When supervising a nuclear power plant, the containment building is crucial. Its functions are guaranteeing structural integrity and avoiding leaks in case of accident. Both events are considered of high risk. Once a given overpressure is registered inside the containment building, three possible outputs are considered: serviceability, breakdown, and collapse. The aim is the study of vulnerability. The vulnerability of the containment building under an overpressure is described by the conditi...

  2. Prediction of failure modes for concrete nuclear-containment buildings

    International Nuclear Information System (INIS)

    Butler, T.A.

    1982-01-01

    The failure modes and associated failure pressures for two common generic types of PWR containments are predicted. One building type is a lightly reinforced, posttensioned structure represented by the Zion nuclear reactor containment. The other is the normally reinforced Indian Point containment. Two-dimensional models of the buildings developed using the finite element method are used to predict the failure modes and failure pressures. Predicted failure modes for both containments involve loss of structural integrity at the intersection of the cylindrical sidewall with the base slab

  3. Biaxial behavior of plain concrete of nuclear containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang-Keun E-mail: sklee0806@bcline.com; Song, Young-Chul; Han, Sang-Hoon

    2004-01-01

    To provide biaxial failure behavior characteristics of concrete of a standard Korean nuclear containment building, the concrete specimens with the dimensions of 200 mmx200 mmx60 mm were tested under different biaxial load combinations. The specimens were subjected to biaxial load combinations covering the three regions of compression-compression, compression-tension, nd tension-tension. To avoid a confining effect due to friction in the boundary surface between the concrete specimen and the loading platen, the loading platens with Teflon pads were used. The principal deformations in the specimens were recorded, and the failure modes along with each stress ratio were examined. Based on the strength data, the biaxial ultimate strength envelopes were developed and the biaxial stress-strain responses in three different biaxial loading regions were plotted. The test results indicated hat the concrete strength under equal biaxial compression, f{sub 1}=f{sub 2}, is higher by about 17% on the average than that under the uniaxial compression and the concrete strength under biaxial tension is almost independent of the stress ratio and is similar to that under the uniaxial tension.

  4. Biaxial behavior of plain concrete of nuclear containment building

    International Nuclear Information System (INIS)

    Lee, Sang-Keun; Song, Young-Chul; Han, Sang-Hoon

    2004-01-01

    To provide biaxial failure behavior characteristics of concrete of a standard Korean nuclear containment building, the concrete specimens with the dimensions of 200 mmx200 mmx60 mm were tested under different biaxial load combinations. The specimens were subjected to biaxial load combinations covering the three regions of compression-compression, compression-tension, nd tension-tension. To avoid a confining effect due to friction in the boundary surface between the concrete specimen and the loading platen, the loading platens with Teflon pads were used. The principal deformations in the specimens were recorded, and the failure modes along with each stress ratio were examined. Based on the strength data, the biaxial ultimate strength envelopes were developed and the biaxial stress-strain responses in three different biaxial loading regions were plotted. The test results indicated hat the concrete strength under equal biaxial compression, f 1 =f 2 , is higher by about 17% on the average than that under the uniaxial compression and the concrete strength under biaxial tension is almost independent of the stress ratio and is similar to that under the uniaxial tension

  5. Requirements for the coatings of a nuclear power plant containment building

    International Nuclear Information System (INIS)

    Orantie, K.; Kuosa, H.; Haekkae-Roennholm, E.

    2001-06-01

    The report presents the criteria for the inside coatings of nuclear power plant containment buildings including: radiation resistance, decontamination, chemical resistance in accident situations and fire resistance

  6. Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

    OpenAIRE

    Shenton B.; Philipose K.

    2011-01-01

    The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in th...

  7. The nonlinear finite element analysis program NUCAS (NUclear Containment Analysis System) for reinforced concrete containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Jin; Lee, Hong Pyo; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The maim goal of this research is to develop a nonlinear finite element analysis program NUCAS to accurately predict global and local failure modes of containment building subjected to internal pressure. In this report, we describe the techniques we developed throught this research. An adequate model to the analysis of containment building such as microscopic material model is adopted and it applied into the development Reissner-Mindlin degenerated shell element. To avoid finite element deficiencies, the substitute strains based on the assumed strain method is used in the shell formulation. Arc-length control method is also adopted to fully trace the peak load-displacement path due to crack formation. In addition, a benchmark test suite is developed to investigate the performance of NUCAS and proposed as the future benchmark tests for nonlinear analysis of reinforced concrete. Finally, the input format of NUCAS and the examples of input/output file are described. 39 refs., 65 figs., 8 tabs. (Author)

  8. Electric cable insulation pyrolysis and ignition resulting from potential hydrogen burn scenarios for nuclear containment buildings

    International Nuclear Information System (INIS)

    Berlad, A.L.; Jaung, R.; Pratt, W.T.

    1982-01-01

    Electric cable insulation in nuclear containment buildings may participate in pyrolysis and combustion processes engendered by hydrogen burn phenomena. This paper examines these pyrolysis/ignition processes of those polymeric materials present in the electric cable insulation and their possible relation to hydrogen burn scenarios

  9. Study of Real Time Location System For Worker in Containment Building at Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. Y.; Kim, G. S. [Samchang Enterprise Company, Ulsan (Korea, Republic of); Kim, H. S. [Ulsan Univ., Ulsan (Korea, Republic of)

    2012-03-15

    Workers are required special management to minimize radiation exposure in nuclear power plant. Especially, there are many limitation in their activities at containment building in nuclear power plant. Test personnel shall administer the workers by tracing the location of them inside containment building in nuclear power plant. They may be exposed to the unnecessary radiation due to a complex and high radiation area in the building. Test personnel needs to manage efficiently for worker's safety and work hours at containment building. Therefore, it is critical for the test personnel to notice the risk to the workers by identifying the location when the workers are facing the dangerous situation on the high area. In this paper, we introduce requirements and design method to develop the one and two dimensional RTLS(Real Time Locating System) by using CSS(Chirp Spread Spectrum) which enables precise location measurement and robust data communication even indoor environment with serious electromagnetic interference caused by complicated structure such as the inside of containment building in the nuclear power plant. In the algorithm to compute the distance, it is suggested to use SDS-TWR(Symmetrical Double-Sided Two-Way Ranging) to solve the issue of indirect routes, and develop the power circuit with 10mW of designing gain for output power to meet the KCC standard in order to increase the raging distance, in addition, communication between Anchor and distance measuring computer shall be designed to increase energy using time of Tags(nodes) by using CAN(Controller Area Network) communication.

  10. Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

    Science.gov (United States)

    Philipose, K.; Shenton, B.

    2011-04-01

    The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

  11. Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

    Directory of Open Access Journals (Sweden)

    Shenton B.

    2011-04-01

    Full Text Available The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA. To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1, Quebec, Canada (250 MWe was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

  12. Assessment of Effective Prestressed Force of Nuclear Containment Building using SI Technique

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. P.; Jang, J. B.; Hwang, K. M.; Song, Y. C. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Bonded tendons have been used in reactor buildings of heavy water reactors and the light water reactors of some nuclear power plants operating in Korea. The assessment of prestressed forces on those bonded tendons is becoming an important issue in assuring their continuous operation beyond their design life. In order to assess the effective prestressed force on the bonded tendon, indirect assessment techniques have been applying to the test beams which were manufactured on construction time. Therefore, this research mainly forced to establish the assessment methodology to measure directly the effective prestressed force on the bonded tendon of containment buildings using System Identification (SI) technique. To accomplish this purpose, simple SI method was proposed and adapted three dimensional finite element analysis of the 1:4 scale prestressed concrete containment vessel (PCCV) tested by Sandia National Laboratory in 2000

  13. Neutron dosimetry inside the containment building of Spanish nuclear power plants with PADC based dosemeters

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Fuste, M.J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Domingo, C., E-mail: carles.domingo@uab.ca [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain); Amgarou, K.; Bouassoule, T.; Castelo, J. [Grup de Fisica de les Radiacions. Departament de Fisica. Edifici C. Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain)

    2009-10-15

    The Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear, CSN) recommends performing neutron individual dose assignments at workplaces based on ambient dose equivalent measurements using area monitors and by estimating the amount of time that workers spend in the different monitored environments. In addition, some Spanish nuclear power plants estimate the neutron dose equivalent using albedo thermoluminescence dosemeters (TLD). In the period 2004-2006, our group, together with other research centers, participated in a project, funded by the CSN, with the support of the Spanish Nuclear Power Plants Association (UNESA), to investigate in situ which could be the best practical procedure for individual neutron dose monitoring in nuclear power plants. As part of this survey, several units of the UAB PADC based neutron dosemeter were exposed, on a methacrylate phantom simulating a human body, at four different places inside the containment building of the Asco I nuclear power plant. The influence of different types of calibration neutron fields is analysed and the dose equivalent for each point is estimated.

  14. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43/sup 0/C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined ..gamma..-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area.

  15. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    International Nuclear Information System (INIS)

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43 0 C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined γ-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area

  16. Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-03-15

    From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

  17. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hong Pyo, E-mail: hplee@kepri.re.k [Nuclear Power Laboratory, Korea Electric Power Research Institute, 103-16 Munji-Dong, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of)

    2011-02-15

    Research highlights: Finite element program with 9-node degenerated shell element was developed. The developed program was mainly forced to analyze nuclear containment building. Concrete material model is adapted Niwa and Yamada failure criteria. The performance of program developed is verified through various numerical examples. The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  18. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    International Nuclear Information System (INIS)

    Lee, Hong Pyo

    2011-01-01

    Research highlights: → Finite element program with 9-node degenerated shell element was developed. → The developed program was mainly forced to analyze nuclear containment building. → Concrete material model is adapted Niwa and Yamada failure criteria. → The performance of program developed is verified through various numerical examples. → The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  19. Contribution to the simulation of the behavior of containment buildings in nuclear power plants

    International Nuclear Information System (INIS)

    Jason, Ludovic

    2016-01-01

    In this document, containment buildings of nuclear power plants in reinforced and prestressed concrete are studied. The mechanical behavior is first investigated. It includes studies at different scales, from the industrial scale at which it is difficult to obtain detailed information about the crack openings, to the structural scale at which it becomes possible to develop more refined approaches (steel-concrete bond, modeling of the different heterogeneities, constitutive models for concrete,...). When it is possible, the developments are based on the combination between modeling, simulation and experiments. A new steel-concrete bond model is especially developed. Its interest, compared to the classical 'perfect bond' hypothesis is particularly discussed. An adaptive condensation technique is finally proposed to bridge the gap between the scale of the containment and the scale of Structural Representative Volumes. As the tightness of the containment buildings is related to the fluid flow through concrete, two approaches are proposed and compared. The first one is based on a chained hydro mechanical simulation and tends to show that the mechanical damage is only influent when it crosses the whole concrete section. In this case, a localized approach is preferred from which a reference crack opening is defined. (author) [fr

  20. Improvement of impact-resistance of a nuclear containment building using fiber reinforced concrete

    International Nuclear Information System (INIS)

    Jeon, Se-Jin; Jin, Byeong-Moo

    2016-01-01

    Highlights: • Impact-resistance of a structure can be improved by fiber reinforced concrete (FRC). • Material modeling of FRC is incorporated into finite element analysis of a structure. • A new index for impact-resistance is proposed based on plastic dissipation energy. • A nuclear power plant made of FRC shows improved resistance against aircraft crashes. - Abstract: Since the act of terrorism that occurred in the USA on September 11, 2001, the protection of nuclear power plants against large commercial aircraft crashes has been an emerging issue. Besides the verification of the safety of nuclear power plants in operation or in design, efficient methods for improving the impact-resistance of these structures have been investigated. Fiber reinforced concrete (FRC) has been generally accepted as an effective material for this purpose. In particular, FRC has been developed to improve the tensile behavior of concrete such as tensile strength, ductility and toughness. One of the main fields of application of FRC can be found in blast-protective or blast-resistant concrete structures. It is expected, therefore, that safety-related structures in a nuclear power plant can also be effectively protected from external blast, aircraft crash, etc. by applying FRC. In order to analytically verify the effect on structural behavior of applying FRC, the particular material properties of FRC should be incorporated into the material modeling of a structural analysis program. This study investigates the mathematical modeling of FRC, which represents various aspects of material behavior. Two numerical examples are provided to show the improved impact-resistance of a nuclear containment building that is expected when applying FRC in comparison with ordinary concrete. The analysis results show that the displacement decreases by 43–67% while the impact-resistance increases by 40–82%, depending on a fiber type.

  1. Improvement of impact-resistance of a nuclear containment building using fiber reinforced concrete

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Se-Jin, E-mail: conc@ajou.ac.kr [Ajou University, 206, World cup-ro, Yeongtong-gu, Suwon-si, Gyeonggi-do 16499 (Korea, Republic of); Jin, Byeong-Moo [DAEWOO E& C, Institute of Construction Technology, 20, Suil-ro 123beon-gil, Jangan-gu, Suwon-si, Gyeonggi-do 16297 (Korea, Republic of)

    2016-08-01

    Highlights: • Impact-resistance of a structure can be improved by fiber reinforced concrete (FRC). • Material modeling of FRC is incorporated into finite element analysis of a structure. • A new index for impact-resistance is proposed based on plastic dissipation energy. • A nuclear power plant made of FRC shows improved resistance against aircraft crashes. - Abstract: Since the act of terrorism that occurred in the USA on September 11, 2001, the protection of nuclear power plants against large commercial aircraft crashes has been an emerging issue. Besides the verification of the safety of nuclear power plants in operation or in design, efficient methods for improving the impact-resistance of these structures have been investigated. Fiber reinforced concrete (FRC) has been generally accepted as an effective material for this purpose. In particular, FRC has been developed to improve the tensile behavior of concrete such as tensile strength, ductility and toughness. One of the main fields of application of FRC can be found in blast-protective or blast-resistant concrete structures. It is expected, therefore, that safety-related structures in a nuclear power plant can also be effectively protected from external blast, aircraft crash, etc. by applying FRC. In order to analytically verify the effect on structural behavior of applying FRC, the particular material properties of FRC should be incorporated into the material modeling of a structural analysis program. This study investigates the mathematical modeling of FRC, which represents various aspects of material behavior. Two numerical examples are provided to show the improved impact-resistance of a nuclear containment building that is expected when applying FRC in comparison with ordinary concrete. The analysis results show that the displacement decreases by 43–67% while the impact-resistance increases by 40–82%, depending on a fiber type.

  2. Floor Response Spectra of Nuclear Containment Building with Soil-Structure Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Choon Gyo; Ryu, Jeong Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    This paper presents a seismic analysis technique for a 3D soil-structure interaction(SSI) system in frequency domain, based on the finite element formulation incorporating frequency-dependent dynamic infinite elements for the far field soil region. Earthquake input motions are regarded as traveling SV-wave which is vertically incident from a far-field soil region. In which, the equivalent earthquake forces in the frequency domain are calculated using the exterior rigid boundary method and the free field response analysis. For the application, floor response spectra analyses for nuclear containment building on a soil medium is carried out, the obtained results are compared with the free field response by other solution.

  3. Fiber reinforced concrete as a material for nuclear reactor containment buildings

    International Nuclear Information System (INIS)

    Mallikarjuna; Banthia, N.; Mindess, S.

    1991-01-01

    The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

  4. GASFLOW: A computational model to analyze accidents in nuclear containment and facility buildings

    International Nuclear Information System (INIS)

    Travis, J.R.; Nichols, B.D.; Wilson, T.L.; Lam, K.L.; Spore, J.W.; Niederauer, G.F.

    1993-01-01

    GASFLOW is a finite-volume computer code that solves the time-dependent, compressible Navier-Stokes equations for multiple gas species. The fluid-dynamics algorithm is coupled to the chemical kinetics of combusting liquids or gases to simulate diffusion or propagating flames in complex geometries of nuclear containment or confinement and facilities' buildings. Fluid turbulence is calculated to enhance the transport and mixing of gases in rooms and volumes that may be connected by a ventilation system. The ventilation system may consist of extensive ductwork, filters, dampers or valves, and fans. Condensation and heat transfer to walls, floors, ceilings, and internal structures are calculated to model the appropriate energy sinks. Solid and liquid aerosol behavior is simulated to give the time and space inventory of radionuclides. The solution procedure of the governing equations is a modified Los Alamos ICE'd-ALE methodology. Complex facilities can be represented by separate computational domains (multiblocks) that communicate through overlapping boundary conditions. The ventilation system is superimposed throughout the multiblock mesh. Gas mixtures and aerosols are transported through the free three-dimensional volumes and the restricted one-dimensional ventilation components as the accident and fluid flow fields evolve. Combustion may occur if sufficient fuel and reactant or oxidizer are present and have an ignition source. Pressure and thermal loads on the building, structural components, and safety-related equipment can be determined for specific accident scenarios. GASFLOW calculations have been compared with large oil-pool fire tests in the 1986 HDR containment test T52.14, which is a 3000-kW fire experiment. The computed results are in good agreement with the observed data

  5. GASFLOW: the theoretical model to analyze accidents in nuclear containments, confinements, and facility buildings

    International Nuclear Information System (INIS)

    Travis, J.R.; Wilson, T.L.

    1993-01-01

    This report documents the governing physical equations for GASFLOW, a finite-volume computer code for solving transient, three-dimensional, compressible, Navier-Stokes equations for multiple gas species. The code is designed to be a best-estimate tool for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments, confinements, and other facility buildings. An analysis with GASFLOW will result in time-dependent gas species concentrations throughout the structure analyzed, and in the event of combustion the pressure and temperature loadings on the walls and internal structures. GASFLOW can model geometrically complex containment systems with multiple compartments and internal structures. It can calculate gas behavior of low-speed buoyancy-driven flows, of diffusion-dominated flows, and during deflagrations. The code can model condensation heat transfer to walls and internal structures by natural and forced convection; chemical kinetics of combustion of hydrogen or hydrocarbon fuels; and fluid turbulence. Heat conduction within walls and structures is considered one-dimensional. 17 refs., 1 fig., 1 tab

  6. Failure modes of a concrete nuclear-containment building subjected to hydrogen detonation

    International Nuclear Information System (INIS)

    Fugelso, L.E.; Butler, T.A.

    1983-01-01

    Calculated response for the Indian Point reactor containment building to static internal pressure and one case of a dynamic pressure representing hydrogen combustion and detonation are presented. Comparison of the potential failure modes is made. 9 figures

  7. XPS investigations of ruthenium deposited onto representative inner surfaces of nuclear reactor containment buildings

    Energy Technology Data Exchange (ETDEWEB)

    Mun, C. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de la Prevention des Accidents Majeurs (DPAM), Centre de Cadarache, BP3-13115 Saint-Paul-lez-Durance (France)]. E-mail: christian.mun@irsn.fr; Ehrhardt, J.J. [Laboratoire de Chimie Physique et Microbiologie pour l' Environnement (LCPME) UMR 7564, CNRS-Nancy University-405, rue de Vandoeuvre 54600 Villers-les-Nancy (France)]. E-mail: ehrhardt@lcpe.cnrs-nancy.fr; Lambert, J. [Laboratoire de Chimie Physique et Microbiologie pour l' Environnement (LCPME) UMR 7564, CNRS-Nancy University-405, rue de Vandoeuvre 54600 Villers-les-Nancy (France); Madic, C. [Commissariat a l' Energie Atomique (CEA), Direction de l' Energie Nucleaire, Centre de Saclay, 91191 Gif-sur-Yvette Cedex (France)]. E-mail: charles.madic@cea.fr

    2007-07-15

    In the case of a hypothetical severe accident in a nuclear power plant, interactions of gaseous RuO{sub 4} with reactor containment building surfaces (stainless steel and epoxy paint) could possibly lead to a black Ru-containing deposit on these surfaces. Some scenarios include the possibility of formation of highly radiotoxic RuO{sub 4}(g) by the interactions of these deposits with the oxidizing medium induced by air radiolysis, in the reactor containment building, and consequently dispersion of this species. Therefore, the accurate determination of the chemical nature of ruthenium in the deposits is of the high importance for safety studies. An experiment was designed to model the interactions of RuO{sub 4}(g) with samples of stainless steel and of steel covered with epoxy paint. Then, these deposits have been carefully characterised by scanning electron microscopy (SEM/EDS), electron probe microanalysis (EPMA) and X-ray photoelectron spectroscopy (XPS). The analysis by XPS of Ru deposits formed by interaction of RuO{sub 4}(g), revealed that the ruthenium is likely to be in the IV oxidation state, as the shapes of the Ru 3d core levels are very similar with those observed on the RuO{sub 2}.xH{sub 2}O reference powder sample. The analysis of O 1s peaks indicates a large component attributed to the hydroxyl functional groups. From these results, it was concluded that Ru was present on the surface of the deposits as an oxyhydroxide of Ru(IV). It has also to be pointed out that the presence of 'pure' RuO{sub 2}, or of a thin layer of RuO{sub 3} or Ru{sub 2}O{sub 5}, coming from the decomposition of RuO{sub 4} on the surface of samples of stainless steel and epoxy paint, could be ruled out. These findings will be used for further investigations of the possible revolatilisation phenomena induced by ozone.

  8. Structural response of nuclear containment shield buildings with unanticipated construction openings

    Science.gov (United States)

    Mac Namara, Sinead Caitriona

    As Nuclear Power Plants age many require steam generator replacement. There is a nickel alloy in the steam generator tubes that is susceptible to stress cracking and although these cracks can be sealed the generator becomes uneconomical without 10%-15% of the tubes. The steam generator in a typical nuclear power plant is housed in the containment structure next to the reactor. The equipment hatch is not big enough to facilitate steam generator replacement, thus construction openings in the dome of the containment structure are required. To date the structural consequences of construction openings in the dome have not been examined. This thesis examines the effects of such openings. The prototype concrete dome is made up of a 2 ft thick dome atop 3 ft thick and 170 ft high cylindrical walls (radius 65.5 ft) with a tension ring 15 ft high and 8 ft thick in between. The dome of the building is cast in two layers; a lower 9 inch layer that serves as the formwork for an upper 15 inch layer. The weight of the dome is carried in axial compression along the hoops and meridians of the dome. The first finite element model uses shell elements and considers two limiting load cases; where the two layers act as one, and where the lower layer carries the weight of both. The openings interrupt the hoops and meridians and the weight of the dome must be redistributed around the openings. Without openings, the stresses due to dead load in the structure are very low when compared to the material strength. The impact of the openings is increased compression stresses near the opening. The maximum stresses are approximately four times larger than in the original structure. These results are confirmed by the second model which is made from layers of solid elements. This model shows a significant difference between the compression on the top surface of the dome, in the affected areas, and that on the bottom surface, leading to shear stresses. These shear stresses are largest around the

  9. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  10. New lumped-mass-stick model based on modal characteristics of structures: development and application to a nuclear containment building

    Science.gov (United States)

    Roh, Hwasung; Lee, Huseok; Lee, Jong Seh

    2013-06-01

    In this study, a new lumped-mass-stick model (LMSM) is developed based on the modal characteristics of a structure such as eigenvalues and eigenvectors. The simplified model, named the "frequency adaptive lumped-massstick model," hasonly a small number of stick elements and nodes to provide the same natural frequencies of the structure and is applied to a nuclear containment building. To investigate the numerical performance of the LMSM, a time history analysis is carried out on both the LMSM and the finite element model (FEM) for a nuclear containment building. A comparison of the results shows that the dynamic responses of the LMSM in terms of displacement and acceleration are almost identical to those of the FEM. In addition, the results in terms of fl oor response spectra at certain elevations are also in good agreement.

  11. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code

    International Nuclear Information System (INIS)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-01-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  12. Elements for computing and forecasting the leakage rate of the inner containment of nuclear reactor buildings

    International Nuclear Information System (INIS)

    Asali, M.; Capra, B.; Mazars, J.; Colliat, J.B.

    2015-01-01

    This study aims at introducing a methodology based on a macro-element discretization to compute and forecast the air leak rate of double-wall reactor buildings during air pressure tests. Assumptions at the basis of a weakly coupled strategy are checked in the case of a typical porous concrete section of an inner containment modeled during a 33 year period including four decennial regulatory pressure tests. However, air leakage due to porosity is only part of in situ measurements. Leakage due to cracking is another part and should be taken into account. A first macro-element is then presented, that superimposes Darcy flow within a porous matrix together with Poiseuille flow within a crack. Those elements are then used in a 3D hydraulic model to compute more accurately the total air leakage rate of considered structures. (authors)

  13. Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

    Directory of Open Access Journals (Sweden)

    Ahmer Ali

    2017-06-01

    Full Text Available Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB under strong short-period ground motions (SPGMs and long-period ground motions (LPGMs. The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

  14. Design response spectra-compliant real and synthetic GMS for seismic analysis of seismically isolated nuclear reactor containment building

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Ahmer [ENVICO Consultants Co. Ltd., Seoul (Korea, Republic of); Abu-Hayah, Nadin; Kim, Doo Kie [Civil and Environmental Engineering, Kunsan National University, Gunsan (Korea, Republic of); Cho, Sung Gook [Innose Tech Co., Ltd., Incheon (Korea, Republic of)

    2017-06-15

    Due to the severe impacts of recent earthquakes, the use of seismic isolation is paramount for the safety of nuclear structures. The diversity observed in seismic events demands ongoing research to analyze the devastating attributes involved, and hence to enhance the sustainability of base-isolated nuclear power plants. This study reports the seismic performance of a seismically-isolated nuclear reactor containment building (NRCB) under strong short-period ground motions (SPGMs) and long-period ground motions (LPGMs). The United States Nuclear Regulatory Commission-based design response spectrum for the seismic design of nuclear power plants is stipulated as the reference spectrum for ground motion selection. Within the period range(s) of interest, the spectral matching of selected records with the target spectrum is ensured using the spectral-compatibility approach. NRC-compliant SPGMs and LPGMs from the mega-thrust Tohoku earthquake are used to obtain the structural response of the base-isolated NRCB. To account for the lack of earthquakes in low-to-moderate seismicity zones and the gap in the artificial synthesis of long-period records, wavelet-decomposition based autoregressive moving average modeling for artificial generation of real ground motions is performed. Based on analysis results from real and simulated SPGMs versus LPGMs, the performance of NRCBs is discussed with suggestions for future research and seismic provisions.

  15. Nuclear reactor building

    International Nuclear Information System (INIS)

    Oshima, Nobuaki.

    1991-01-01

    The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

  16. Excessive leakage measurement using pressure decay method in containment building local leakage rate test at nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Kyu; Kim, Chang Soo; Kim, Wang Bae [KHNP, Central Research Institute, Daejeon (Korea, Republic of)

    2016-06-15

    There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

  17. Comparison of test and earthquake response modeling of a nuclear power plant containment building

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, M.G.; Kot, C.A.; Hsieh, B.J.

    1985-01-01

    The reactor building of a BWR plant was subjected to dynamic testing, a minor earthquake, and a strong earthquake at different times. Analytical models simulating each of these events were devised by previous investigators. A comparison of the characteristics of these models is made in this paper. The different modeling assumptions involved in the different simulation analyses restrict the validity of the models for general use and also narrow the comparison down to only a few modes. The dynamic tests successfully identified the first mode of the soil-structure system.

  18. Comparison of test and earthquake response modeling of a nuclear power plant containment building

    International Nuclear Information System (INIS)

    Srinivasan, M.G.; Kot, C.A.; Hsieh, B.J.

    1985-01-01

    The reactor building of a BWR plant was subjected to dynamic testing, a minor earthquake, and a strong earthquake at different times. Analytical models simulating each of these events were devised by previous investigators. A comparison of the characteristics of these models is made in this paper. The different modeling assumptions involved in the different simulation analyses restrict the validity of the models for general use and also narrow the comparison down to only a few modes. The dynamic tests successfully identified the first mode of the soil-structure system

  19. The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

    International Nuclear Information System (INIS)

    Rouelle, P.; Roy, F.

    1998-01-01

    The Electricite de France's N4 CHOOZ B nuclear power plant, two units of the world's largest PWR model (1450 Mwe each), has earned the Electric Power International's 1997 Powerplant Award. This lead NPP for EDF's N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building's inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter

  20. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  1. Ageing management of CANDUtm concrete containment buildings

    International Nuclear Information System (INIS)

    Philipose, K.E.; Gregor, F.E.

    2009-01-01

    The containment system in a Nuclear Power Plant (NPP) provides the final physical barrier against release of radioactive materials to the external environment. Even though there are different physical configurations to meet this fundamental safety function in various reactor types, a common feature is the use of a thick-walled concrete structure as part of the containment system commonly referred to as 'Concrete Containment Building'. In order for the concrete containment buildings to continue to provide the required safety function, it has to maintain its structural integrity. As well, its leak rates under test pressures must be maintained below acceptable limits. As some of the containment buildings of the CANDU nuclear power plants are approaching their fourth decade of successful operation, questions regarding the impact of ageing on their ultimate useful service life emerge. Ageing Management has become the tool for addressing those questions. In this paper, the ageing and ageing management of the CANDU concrete containments are discussed, including the specific programs being implemented to monitor and trend the ageing conditions. Specifically, the usefulness of the embedded strain gauges as a tool for the assessment of the condition of the containment concrete structure is discussed. Some of the operational and test data accumulated over the last 30 years have been evaluated and trended to provide some results and conclusions regarding the satisfactory long-term behaviour of the concrete containment buildings. (authors)

  2. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

    The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

  3. A Monte Carlo program to calculate the exposure rate from airborne radioactive gases inside a nuclear reactor containment building.

    Science.gov (United States)

    Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W

    1986-12-01

    A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.

  4. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  5. Assessment and management of ageing of major nuclear power plant components important to safety: Concrete containment buildings

    International Nuclear Information System (INIS)

    1998-06-01

    The report presents the results of the Co-ordinated Research Programme (CRP) on the Management of Ageing of Concrete Containment Buildings (CCBs) addressing current practices and techniques for assessing fitness-for-service and the inspection, monitoring and mitigation of ageing degradation of selected components of CANDU reactor, BWR reactor, PWR reactor and WWER plants. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs and also to provide a common technical basis for dialogue between plant operators and regulators when dealing with age-related licensing issues

  6. Nuclear reactor container

    International Nuclear Information System (INIS)

    Yamaki, Rika; Kawabe, Ryuhei.

    1989-01-01

    A venturi scrubber is connected to a nuclear reactor container. Gases containing radioactive aerosols in the container are introduced into the venturi scrubber in the form of a high speed stream under the pressure of the container. The radioactive aerosols are captured by inertia collision due to the velocity difference between the high speed gas stream and water droplets. In the case of the present invention, since the high pressure of the reactor container generated upon accident is utilized, compressor, etc. is no more required, thereby enabling to reduce the size of the aerosol removing device. Further, since no external power is used, the radioactive aerosols can be removed with no starting failure upon accidents. (T.M.)

  7. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

  8. Multi-scale approach of the mechanical behaviour of reinforced concrete structures - Application to nuclear plant containment buildings

    International Nuclear Information System (INIS)

    David, M.

    2012-01-01

    This thesis develops a multi-scale strategy to describe the mechanical behaviour of steel reinforcements and prestressing tendons in a reinforced concrete structure. This strategy is declined in several steps, which allow gradual integration of new physical phenomena. The first asymptotic model represents the effective elastic behaviour of heterogeneities periodically distributed on a surface. It combines an elastic interface behaviour and a membrane behaviour. A second asymptotic model then focuses on the behaviour of rigid fibers distributed on a surface, which may slide with respect to the surrounding volume. These models induce less stress concentrations than the usual truss models. They are implemented in the finite element code Code-Aster, and validated with respect to reference three-dimensional simulations. Their interaction with a macroscopic crack is studied. Finally, this strategy allows the modeling of experimental tests carried out on a portion of a containment building in real scale. (author)

  9. In-depth analysis of eight criteria for integrated leakage rate tests for nuclear power plant containment buildings

    International Nuclear Information System (INIS)

    Wagner, W.T.; Langan, J.P.; Norris, W.F.; Lurie, D.

    1989-01-01

    A U.S. Nuclear Regulatory Commission (NRC) Small Business Innovation research (SBIR) Contract investigated ten integrated leakage rate test (ILRT) analysis models which have been proposed for evaluation of ILRT data. This contract involved in-depth analysis of two ILRTs with data collected at accelerated rates and 80 conventional ILRTs with data collected at a frequency between 10-15 minutes. All ten methods were applied to all data. The study considered the appropriateness of each method to analyze containment data (air mass versus time), the influence of data collection frequency on ILRT duration, and the influence of collection frequency on each method. The study is described in the paper. Results are presented

  10. Building of nuclear power plant

    International Nuclear Information System (INIS)

    Saito, Takashi.

    1997-01-01

    A first nuclear plant and a second nuclear power plant are disposed in adjacent with each other in a building for a nuclear reactor. A reactor container is disposed in each of the plants, and each reactor container is surrounded by a second containing facility. A repairing chamber capable of communicating with the secondary containing facilities for both of the secondary containing facilities is disposed being in contact with the second containing facility of each plant for repairing control rod driving mechanisms or reactor incorporated-type recycling pumps. Namely, the repairing chamber is in adjacent with the reactor containers of both plants, and situated between both of the plants as a repairing chamber to be used in common for both plants. Air tight inlet/exit doors are formed to the inlets/exits of both plants of the repairing chamber. Space for the repairing chamber can be reduced to about one half compared with a case where the repairing chamber is formed independently on each plant. (I.N.)

  11. Nuclear reactor container

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1989-01-01

    Aerosol filters considered so far for nuclear reactor containers in conventional BWR type nuclear power plants make the facility larger and involve a risk of clogging. In view of the above, in the present invention, the diameter of a flow channel of gases entering from a bent pipe to a suppression pool is made smaller thereby decreasing the diameter of gas bubbles in the supperssional pool. Since this reduces the force of surface tension, the diameter of resulted gas bubbles is made remarkably smaller as compared with the case where the gases are released from the lower end of the bent pipe. Since the absorption velocity of bubble-entrained aerosols into water is in proportion to the square of the bubble diameter, the absorption efficiency can be increased remarkably by reducing the diameter of the gas bubbles. Accordingly, it is possible to improve the efficiency of eliminating radioactivity of released gases. (K.M.)

  12. Analysis of Gamma Dose Rate Caused by Corrosion Products inside the Containment Building of Yonngwang Nuclear Power Plant Unit 3 During Shutdown Period

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Wi Ho; Kim, Jae Cheon; Kim, Soon Young; Kim, Jong Kyung [Hanyang Univ., Seoul (Korea, Republic of)

    2005-07-01

    Occupational radiation exposure(ORE) of nuclear power plant(NPP) workers mainly occurs during the shutdown period. Major radioactive sources are the corrosion products released from the reactor coolant system(RCS). The corrosion products consist of circulating crud and deposited crud. Major radioactive corrosion products, {sup 58}Co and {sup 60}Co, are known to contribute approximately more than 70% of the total ORE. In this study, the corrosion products regarding cobalt were evaluated during the shutdown period, and gamma dose rates caused by them were calculated at the main working area inside the containment building of the Yonggwang NPP Unit 3.

  13. Nuclear reactor container

    International Nuclear Information System (INIS)

    Moriyama, Takeo; Ochiai, Kanehiro; Niino, Tsuyoshi; Kodama, Toyokazu; Hirako, Shizuka.

    1988-01-01

    Purpose: To obtain structures suitable to a container structures for nuclear power plants used in those districts where earthquakes occur frequently, in which no local stresses are caused to the fundamental base portions and the workability for the fundamental structures is improved. Constitution: Basic stabilizers are attached to a nuclear reactor container (PCV) and a basic concrete recess for receiving a basic stabilizer is disposed in basic concretes. A top stabilizer is joined and fixed to a top stabilizer receiving plate at the inside of an outer shielding wall. On the other hand, a PCV top recess for conducting the load of PCV to the top stabilizer is attached to the top of the PCV. By disposing stabilizer structures allowing miner displacements at the two points, that is, the top and the lowermost portion of the PCV, no local stress concentrations can be generated to the extension on the axial direction of components due to the inner pressure of the PCV and to the horizontal load applied to the upper portion of the PCV upon earthquakes. (Yoshino, Y.)

  14. Feasibility study of the IE-SASW method for nondestructive evaluation of containment building structures in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.S. E-mail: dskim@kaist.ac.kr; Kim, H.W. E-mail: hwk@kaist.ac.kr; Seo, W.S.; Choi, K.C.; Woo, S.K

    2003-02-01

    The IE-SASW method, a combination of impact-echo (IE) acoustics with spectral analysis of surface waves (SASW), is proposed as a newly developed nondestructive testing method in concrete structures. This feasibility study examines the IE technique and uses elastic P-wave velocity data as measured from the SASW method on concrete members in nuclear power plant containment structures. It was shown that both the thickness of the concrete specimens used in this study and the depth of the introduced defects (i.e. voids) could be identified by the IE-SASW method. In contrast, the reinforced steel bar itself could not be identified by the IE-SASW method. Additionally, GPR (ground penetrating radar) techniques were used to examine the same specimens in order to establish some level of performance and reliability to compare with the performance of the IE-SASW method. The GPR method provides an objective and reliable image corresponding to the reinforced steel bars. The experimental studies show that it is more feasible to use the IE-SASW method rather than GPR to detect voids that were positioned beneath the steel reinforcing bars in the concrete specimens.

  15. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fukui, Tooru; Murase, Michio; Kataoka, Yoshiyuki; Hidaka, Masataka; Sumita, Isao; Tominaga, Kenji.

    1992-01-01

    In a nuclear reactor container, a chamber in communication with a wet well of a pressure suppression chamber is disposed and situated to such a position that the temperature is lower than a chamber containing pool water upon occurrence of loss of coolant accident. In addition, the inner surface of the pressure suppression chamber is constituted with steel walls in contact with pool water, and an outer circumferential pool is disposed at the outer circumferential surface thereof. Further, a circulation channel is disposed, and a water intake port is disposed at a position higher than an exit to the pool water, and a water discharge port is opened in the pool water at a position lower than the exit to the pool water. With such a constitution, the allowable temperature of the pressure suppression pool water can be elevated to a saturated steam temperature corresponding to the resistant pressure of the container, so that the temperature difference between the pressure suppression pool and the outer side thereof is increased by so much, to improve thermal radiation performance. Accordingly, it can be utilized as a pressure suppression means for a plant of greater power. Further, thermal conduction efficiency from the pool water region of the pressure suppression chamber to the outer circumferential pool water is improved, or thermal radiation area is enlarged due to the circulation channel, to improve the heat radiation performance. (N.H.)

  16. Nuclear reactor container

    International Nuclear Information System (INIS)

    Shioiri, Akio.

    1992-01-01

    In a nuclear reactor container, a vent tube communication port is disposed to a pressure suppression pool at a position higher than the pool water therein for communication with an upper dry well, and the upper end opening of a dry well communication pipe is disposed at a position higher than the communication port. When condensate return pipeline is ruptured in the upper dry well, water in a water source pool is injected to the pressure vessel and partially discharged out of the ruptured port and a depressurization valve connected to the pressure vessel to the inside of the upper dry well. The discharged water stays in the upper dry well and, when the water level reaches the height of the vent tube communication port, it flows into the pressure suppression pool. Even in a state that the entire amount of water in the water source pool is supplied, since water does not reach the upper opening port of the dry well communication pipe, water does not flow into a lower dry well. Accordingly, the motor of a control rod drives disposed in the lower dry well can be prevented from submerging. The reactor core can be cooled more reliably, to improve the reliability of the pressure suppression function. (N.H.)

  17. Measurement of gamma ray flux within the containment building at the first unit of Kori nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. W.; Kim, K. D.; Yoon, C. H.; Han, J. M.; Hu, Y. H. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

    2004-07-01

    To evaluate gamma ray dose response of GM counter being used for monitoring of gamma ray field in nuclear power plants, gamma ray energy spectra and fluxes were obtained for three positions at the unit 1 of the Kori nuclear power station. By applying the response values of Eberline's E112B survey meter to the results, the doses represented on the survey meter were overestimated from 1.31 to 1.37 times when compared to the real doses for these three positions.

  18. Composite containment for nuclear power

    International Nuclear Information System (INIS)

    Harstead, G.A.; Soeoet, O.

    1977-01-01

    Fundamentally, a nuclear reactor containment structure provides three major functions; namely, (1), to withstand loads due to pressure and temperature increase due to Design Basis Accident (DBA) (2), to withstand environmental loads such as seismic, tornado and normal loads, and (3) act as a radiation shield. Conventional design practise is to employ either a steel vessel and concrete shield building or a steel lined concrete structure. This paper deals with a new concept in which a steel liner is employed which carries much of the primary membrane loads. This type of structure is similar in some aspects to the previously described systems: a) A mat, lined with a thin plate on its top surface, is similar to concrete containment. b) A cylinder and hemispherical dome, made up of steel plate and concrete, is about 2.5 feet thick (the minimum required for radiation shielding). Although the steel plate and concrete are in contact, as in concrete containment, the steel plate in composite containment is much thicker than the liner. There are two main advantages over present practise; namely reduction of materials and therefore reduced capital cost and even more significantly a shortened construction schedule which will permit more flexibility in overall plant construction schedule and will benefit the cash flow situation. (Auth.)

  19. Nuclear power. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, W.C.

    1983-01-01

    Lay language brings an understanding of nuclear technology and nuclear politics to the non-specialist reader. The author notes that there has been little change in the technology during the four decades of the nuclear age, but mankind has still to learn how to live with it. Part One explains how reactors work, identifies different reactor types, and describes the fuel cycle. Part two follows research developments during the pre-Manhatten Project days, the war effort, and the decision to pursue commercial nuclear power. He traces the development of policies to secure fission materials and international efforts to prevent the proliferation of weapons grade material and the safe handling of radioactive wastes on a global as well as national scale. There are four appendices, including an annotated reference to other publications. 9 figures.

  20. Principles of building and assembly technology of containment from steel structural blocks for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Eichstedt, J.; Friedrich, F.

    1983-01-01

    This technology is being developed in cooperation between the USSR and the GDR. The cylindrical part of the containment consists of prefabricated double-sided steel blocks with inner reinforcement. The steel plates in a thickness of 20 mm provide casing and secure tightness. Blocks with one steel wall are used for the construction of the cupola. The outer slabs are assembled subsequently. The methods of assembly, concreting and quality assurance are described. (Ha)

  1. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  2. Development of 3D models of buildings for containment of the nuclear power plant of Almaraz and of the Trillo Nuclear with the GOTHIC 8.0 code; Desarrollo de modelos 3D de los edificios de conten cion de la Central Nuclear de Almaraz y de la Central Nuclear de Trillo con el codigo GOTHIC 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, G.; Bocanegra Melian, R.; Fernandez Cosils, K.; Barreira Pereira, P.; Rey Peinado, L.; Posada Barral, J. M.

    2014-07-01

    The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)

  3. Nuclear reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Murase, Michio; Fujii, Tadashi; Susuki, Akira.

    1994-01-01

    A wet well space above a pressure suppression pool is divided into a first wet well on the side in contact with the pressure suppression pool and a second wet well on the side not in contact with the pool. Cooling water is contained in the second wet well and it is in communication with the first wet well by pipelines. Since steams flown into the second well are condensed in the cooling water, they continuously transfer from the first wet well to the second wet well, thereby capable of eliminating the effects of incondensible gases in the first wet well. With such procedures, the effect of the incondensible gases can be eliminated even without cooling from the outside of the reactor. Heat accumulation can be increased in a container of any material, so that thermal load on cooling circuits for removing after-heat can be mitigated. (T.M.)

  4. Nuclear steam system containment

    International Nuclear Information System (INIS)

    Jabsen, F.

    1980-01-01

    An improved containment used for radiation shielding and pressure suppression comprising a dry well includes a pressure vessel, a plurality of concentric wall means, said plurality of concentric wall means defining at least three annular regions about said dry well. A first annular region provides the containment used for radiation shielding, a second annular region is substantially dry, a third annular region provides a wet well for relieving fluid pressure released from the pressure vessel into the dry well. Pipe connection means extend in the wet well from the dry well, a pool of liquid is disposed to partially fill said third annular region, the upper end portion of the second and third annular regions having an enclosure, and a plurality of baffle plates extending vertically downward from said enclosure in said third annular region into said pool of liquid so as to circumferentially divide the upper portion of said third annular region into a plurality of circumferential upper portions

  5. Building a secondary containment system

    Energy Technology Data Exchange (ETDEWEB)

    Broder, M.F.

    1994-10-01

    Retail fertilizer and pesticide dealers across the United States are installing secondary containment at their facilities or are seriously considering it. Much of this work is in response to new state regulations; however, many dealers not facing new regulations are upgrading their facilities to reduce their liability, lower their insurance costs, or comply with anticipated regulations. The Tennessee Valley Authority`s (TVA) National Fertilizer and Environmental Research Center (NFERC) has assisted dealers in 22 states in retrofitting containment to their facilities. Simultaneous improvements in the operational efficiency of the facilities have been achieved at many of the sites. This paper is based on experience gained in that work and details the rationale used in planning secondary containment and facility modifications.

  6. Mobile nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

    1978-01-01

    A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

  7. Shipping container for nuclear fuels

    International Nuclear Information System (INIS)

    Housholder, W.R.; Greer, N.L.

    1976-01-01

    A container for nuclear materials is described wherein a specially and uniquely constructed pressure vessel and gamma shield assembly for holding the nuclear materials is provided in a housing, and wherein a positioning means extends between the housing and the assembly for spacing the same, insulation in the housing essentially filling the space between the assembly and housing, the insulation comprising beads, globules or the like of water encapsulated in plastic and which, in one important embodiment, contains neutron absorbing matter

  8. Container for nuclear fuel powders

    International Nuclear Information System (INIS)

    Etheredge, B.F.; Larson, R.I.

    1982-01-01

    A critically safe container is disclosed for the storage and rapid discharge of enriched nuclear fuel material in powder form is disclosed. The container has a hollow, slab-shaped container body that has one critically safe dimension. A powder inlet is provided on one side wall of the body adjacent to a corner thereof and a powder discharge port is provided at another corner of the body approximately diagonal the powder inlet. Gas plenum for moving the powder during discharge are located along the side walls of the container adjacent the discharge port

  9. Glass containing radioactive nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1985-01-01

    Lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level-radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800 C, since they exhibit very low melt viscosities in the 800 to 1050 C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550 C and are not adversely affected by large doses of gamma radiation in H 2 O at 135 C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms. (author)

  10. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  11. Pressure suppression device for nuclear reactor building

    International Nuclear Information System (INIS)

    Ikegame, Noboru.

    1992-01-01

    In a nuclear reactor building, there are disposed cooling coils connected to an air supply duct at the outside of the building, an air supply blower, an air supply duct having the top end opened, an exhaustion duct having the top end opened and a bypassing pipeline interposed between the exhaustion duct and the air supply duct on the side of the inlet of the cooling coils. In the reactor building, when a radioactive material leakage accident should occur, an isolation valve is closed to isolate the building from the outside. Further, bypassing isolation valve is opened to form a closed cooling circuit by the cooling coils, the air supply blower and the air supply duct, the exhaustion duct and the bypassing pipeline in the reactor building. With such a constitution, since air as the atmosphere in the building is circulated through the closed cooling circuit and cooled by the cooling coils, the temperature is not elevated. Accordingly, since the pressure elevation of the atmosphere in the building is suppressed, the atmosphere containing radioactive materials do not flow out of the building. (I.N.)

  12. Nuclear power plant containment construction

    International Nuclear Information System (INIS)

    Schabert, H.P.; Danisch, R.; Strickroth, E.

    1975-01-01

    The Nuclear Power Plant Containment Construction includes the spherical steel safety enclosure for the reactor and the equipment associated with the reactor and requiring this type of enclosure. This steel enclosure is externally structurally protected against accident by a concrete construction providing a foundation for the steel enclosure and having a cylindrical wall and a hemispherical dome, these parts being dimensioned to form an annular space surrounding the spherical steel enclosure, the latter and the concrete construction heretofore being concentrically arranged with respect to each other. In the disclosed construction the two parts are arranged with their vertical axis horizontally offset from each other so that opposite to the offsetting direction of the concrete construction a relatively large space is formed in the now asymmetrical annular space in which reactor auxiliary equipment not requiring enclosure by the steel containment vessel or safety enclosure, may be located outside of the steel containment vessel and inside of the concrete construction where it is structurally protected by the latter

  13. Integrated engineering system for nuclear facilities building

    International Nuclear Information System (INIS)

    Tomura, H.; Miyamoto, A.; Futami, F.; Yasuda, S.; Ohtomo, T.

    1995-01-01

    In the construction of buildings for nuclear facilities in Japan, construction companies are generally in charge of the building engineering work, coordinating with plant engineering. An integrated system for buildings (PROMOTE: PROductive MOdeling system for Total nuclear Engineering) described here is a building engineering system including the entire life cycle of buildings for nuclear facilities. A Three-dimensional (3D) building model (PRO-model) is to be in the core of the system (PROMOTE). Data sharing in the PROMOTE is also done with plant engineering systems. By providing these basic technical foundations, PROMOTE is oriented toward offering rational, highquality engineering for the projects. The aim of the system is to provide a technical foundation in building engineering. This paper discusses the characteristics of buildings for nuclear facilities and the outline of the PROMOTE. (author)

  14. Present status of nuclear containments in Korea

    International Nuclear Information System (INIS)

    Park, Jihong; Hong, Jaekeun; Lee, Byunghoon; Son, Youngho

    2007-01-01

    Since the first nuclear power plant in Korea, Kori unit no.1, was started in commercial service in 1978, 20 units including Kori unit no.1 have been operated and maintained until now in Korea. Recently several units were started to be constructed and also, additionally more than 4 units were planned to be constructed in the near future. The importance of nuclear containments has been always one of the hottest issues for the safety and protection of nuclear power plants until now. At the beginning of nuclear power plants construction in Korea, several typed nuclear containment systems were adopted. For those reasons, various codes, standards, and inspection technologies are applied to nuclear containment systems differently. In this study, the status of inservice inspection performed for the safety and maintenance of nuclear containments in Korea was researched. Overall nuclear containment systems and inspections performed up to recently in Korea including trends, inspection items, periods, and regulations were described briefly. (author)

  15. Long-term performance of structures comprising nuclear power plants PART 1: Deterioration assessment of nuclear power station buildings PART 2: Long-term stability and the leak-tightness of reactor containments

    International Nuclear Information System (INIS)

    Pocock, D.C.; Worthington, J.C.; Oberpichler, R.; Van Exel, H.; Beukelmann, D.; Huth, R.; Rose, B.

    1990-01-01

    The objective of this research was to study the long-term performance of structures comprising nuclear power plants. The time period of interest for this study is 140 years (this figure is based on maximum periods of 40 years for operation and 100 years of storage). It was divided in two parts: - the first based on four UK nuclear power plants examine the principle deterioration mechanism of reinforced structure which is chloride ingress and carbonation penetration - the second based on 2 German nuclear power plants examine the long term behaviour of reinforced and prestressed concrete and also the corrosion of steel containments with particular reference on plastic seals and potential risk areas

  16. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  17. Aircraft Impact Assessment of APR1400 Reactor Containment Building

    International Nuclear Information System (INIS)

    Moon, Il Hwan; Kim, Do Yeon; Kim, Jae Hee; Kim, Sang Yun

    2011-01-01

    The implementation of a protection to withstand aircraft impact on safety-related structures and systems is basically based on a probabilistic evaluation for each site, if the licensing body doesn't require a deterministic approach. Existing nuclear power plants in Korea were designed based on the probabilistic approach, and the aircraft impact hazard remained less than a probability of 10 -7 . However, a man-made aircraft impact have been considered as a possible external accident for the nuclear power plant. New plant designs that are to be constructed in the U.S. after July 2009 must consider the effect of impact from a large commercial aircraft according to the requirements of 10 CFR 50.150. Especially, Reactor Containment Building (RCB) housing the safety-related equipment and fuels should be protected safely against aircraft crash without perforation and scabbing failure of external wall. APR1400 RCB is constructed as a prestressed concrete containment vessel (PCCV) which is surrounded by the auxiliary building housing additional safety-related equipment and other systems. In this study, the aircraft impact analyses for the RCB are carried out using Riera forcing function and aircraft model. Considered external wall thickness is 4 ft 6 in. for the cylindrical wall and 4 ft for the dome. Actual strengths of concrete and steel are considered as the material properties. For these analyses, the dynamic increment factor and concrete aging effect are considered in accordance with NEI 07-13(2011)

  18. Building nuclear skills in Australia

    International Nuclear Information System (INIS)

    Cameron, R.

    2007-01-01

    Demand for nuclear skills in Australia has traditionally been met by recruitment but as the nuclear industry grows worldwide, such skills are in demand. This paper discusses he likely numbers of skilled people needed for a nuclear industry in Australia and what initiatives have been, or could be in, taken to address the needs

  19. Hydrogen behavior at a subcomparment in the containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, U. J.; Park, G. C. [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    For hydrogen management in severe accidents with degraded nuclear core of PWR`s, several experiments have been performed in the SNU hydrogen mixing facility. The objectives are understanding the extent of hydrogen mixing and analyzing the effects of factors which dominate uniform or non-uniform mixing at compartments in the containment building. The facility represents on a 1/11th linearly scaled model of the YGN unit 3 and 4, hydrogen was simulated by helium. Because there are the gaps between safety injection tank and compartment layers in the containment, the test facility was constructed in three dimensional mode for analyzing of mixture behaviour through the gaps. >From the experimental results we could conclude that overall hydrogen concentration distributed uniformly in the free volume of the test compartment, but fluctuated in the gaps. This paper is focused on experimental result from several experiments. 5 refs., 4 figs. (Author)

  20. Hydrogen behavior at a subcomparment in the containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, U J; Park, G C [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    For hydrogen management in severe accidents with degraded nuclear core of PWR`s, several experiments have been performed in the SNU hydrogen mixing facility. The objectives are understanding the extent of hydrogen mixing and analyzing the effects of factors which dominate uniform or non-uniform mixing at compartments in the containment building. The facility represents on a 1/11th linearly scaled model of the YGN unit 3 and 4, hydrogen was simulated by helium. Because there are the gaps between safety injection tank and compartment layers in the containment, the test facility was constructed in three dimensional mode for analyzing of mixture behaviour through the gaps. >From the experimental results we could conclude that overall hydrogen concentration distributed uniformly in the free volume of the test compartment, but fluctuated in the gaps. This paper is focused on experimental result from several experiments. 5 refs., 4 figs. (Author)

  1. Nuclear buildings and sustainable development

    International Nuclear Information System (INIS)

    Gomah, A.M.H

    2009-01-01

    The main proposal of this thesis based on some practical notes and the theoretical readings, the mathematical equations which led to existing a shared relationship between the nuclear institutions and the economical development with preserving the environment and its recourses which achieves the concept of the sustainable development. The thesis aims also at recognizing the most important characteristics of the nuclear institutions , as the study interests in understanding how the nuclear energy can be distinguished from the other energy resources. Furthermore, the study in its intellectual framework interests in comparing a number of the nuclear institutions that the study finds them related to the research topic and assists in achieving the study goals, which represent in the environmental evaluation of the nuclear institutions inside its biological surroundings. The study consists of four main chapters in addition to the introduction and the conclusion as follows: The first chapter: Recognizing the nuclear institutions and their effect on the environment. The second chapter: Recognizing planning and generalizing the nuclear institutions. The third chapter: Recognizing the limits and standards of the planning and the designing of a nuclear institution. The fourth chapter: The nuclear institutions inside the suburban places.

  2. Physico-chemical aspects of radionuclide removal under accident conditions in nuclear power plants by means of containment-building spray systems

    International Nuclear Information System (INIS)

    Alm, M.

    1976-01-01

    A survey is given on industrial spray solutions, their thermal and radiation stability, drop effects, and the corrosion behaviour of reactor and containment materials. From the hitherto known spray experiments it may be concluded that (1) spray solutions can be used for the effective decontamination of the containment atmosphere in the event of a loss-of-coolant accident, (2) the spray efficiency for the removal of gaseous and volatile fission products can be assessed by means of simplified model considerations, (3) further work is necessary to optimize the technology of the spray process. (author)

  3. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  4. Development of safety enhancement technology of containment building

    International Nuclear Information System (INIS)

    Seo, Jeong Moon; Choun, Y. S.; Choi, I. K.

    2002-04-01

    This study consists of four research areas, (1) Seismic safety assessment, (2) Aging assessment of a containment building, (3) Prediction of long-term behavior and analysis of a containment building, (4) Performance verification of a containment building. In the seismic safety assessment area, responses of a containment building were monitored and the analysis method was verified. Also performed are the identification of earthquake characteristics and improvement of the seismic fragility analysis method. In the area of aging assessment of a containment building, we developed aging management code SLMS and database. Aging tests were performed for containment building materials and aging models were developed. Techniques for investigation, detection, and evaluation of aging were developed. In the area of prediction of long-term behavior and analysis of a containment building, we developed a non-linear structural analysis code NUCAS and material models. In the area of performance verification of a containment building, we analyzed the crack behavior of a containment wall and the behavior of the containment under internal pressure. We also improved the ISI methods for prestressed containment

  5. Nuclear Buildings And Sustainable Development

    International Nuclear Information System (INIS)

    Gomah, A.El.M.H.

    2009-01-01

    The main proposal of this Thesis based on some practical notes and the theoretical readings, the mathematical equations which led to existing a shared relationship between the nuclear institutions and the economical development with preserving the environment and its recourses which achieves the concept of the sustainable development. The Thesis aims also at recognizing the most important characteristics of the nuclear institutions, as the study interests in understanding how the nuclear energy can be distinguished from the other energy resources. Furthermore, the study in its intellectual framework interests in comparing a number of the nuclear institutions that the study finds them related to the research topic and assists in achieving the study goals ,which represent in the environmental evaluation of the nuclear institutions inside its biological surroundings. The study consists of four main chapters in addition to the introduction and the conclusion as follows: The first chapter: recognizing the nuclear institutions and their effect on the environment. This chapter includes studying the characteristics of the nuclear institutions in the frame of its existence in the atmospheric surroundings and this chapter includes: 1- The kinds of the nuclear institutions, the troubles and incidents resulting in them and comparing between it and the study of the nuclear fuel. 2- The economical importance of the nuclear institutions and participating it in the process of developing. The role of the agency of preserving the environment and the extent of its ability to deal with the nuclear incidents and training and guiding the inhabitants how to deal with these incidents.The second chapter: recognizing planning and generalizing the nuclear institutions.This chapter handles by the study and analysis the nature of the nuclear institutions and the development in their designs according to the development in the designs of the nuclear institutions and this chapter includes:1- The

  6. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  7. Building method for nuclear power plant

    International Nuclear Information System (INIS)

    Iba, Tsutomu.

    1993-01-01

    Outer shielding buildings and inner concretes are dismantled and removed. In this case, foundations are left as they are. Then, new outer shielding buildings and inner concretes are constructed on the old foundations by way of rubber a laminates. The outer shielding buildings and the inner concretes supported by the rubber laminates have an earthquake-proof structure and vibrations upon occurrence of earthquakes can be moderated by the rubber laminates. Therefore, it is not required to make the outer shielding buildings so strong, and the weight of them can be reduced. Accordingly, the weight of the buildings is reduced compared with the buildings before dismantling, so that they can be sufficiently supported by the old foundations. This enables to build a nuclear power plant without ensuring new construction sites. (I.N.)

  8. Detonability of containment building atmospheres during core-meltdown accidents

    International Nuclear Information System (INIS)

    Jaung, R.; Berlad, L.; Pratt, W.

    1983-01-01

    During Core-Meltdown Accidents in Light Water Reactors, significant quantities of combustible gases could be released to the containment building. The highest possible peak pressure fields that may occur through combustion processes are associated with detonation phenomena. Accordingly, it is necessary to understand and identify the possible ways in which detonations may or may not occur. Although no comprehensive theory of detonation is currently available, there are useful guidelines, which can be derived from current theoretical concepts and the body of experimental data. This paper examines these guidelines and indicates how they may be used to evaluate the possible occurrence of detonation-related combustion processes. In particular, this study identifies three features that an initiation source must achieve if it is to ultimately result in a stable detonation. One of these features requires post-shock initial conditions that lead to very short ignition delays. This concept is used to examine the possibility of achieving quasi-steady detonation phenomena in nuclear reactor containment buildings during postulated core-melt accidents

  9. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  10. Darlington GS vacuum building - containment shell

    International Nuclear Information System (INIS)

    Huterer, J.; Ha, E.C.; Brown, D.G.; Cheng, P.C.

    1985-01-01

    The paper describes the consequences of new design requirements for the Darlington vacuum building on its structural configuration, analytical and reinforcing steel layout. Attention focuses on the ring girder where the juncture of dome and perimeter wall produces a complex post-tensioning layout, and attendant difficulties in design and construction. At the wall base, full fixity imposes large local stresses. Long-term, shrinkage and creep, and temperature effects become significant. A research program and in-house analytical procedure established time-dependent concrete behaviour and corresponding wall-sectional stresses. The outcome is examined in terms of reinforcement, temperature controls, and wall liner requirements. (orig.)

  11. Aircraft transporting container for nuclear fuel

    International Nuclear Information System (INIS)

    Kurakami, Jun-ichi; Kubo, Minoru.

    1991-01-01

    The present invention concerns an air craft transporting container for nuclear fuels. A sealing container that seals a nuclear fuel container and constitutes a sealed boundary for the transporting container is incorporated in an inner container. Shock absorbers are filled for absorbing impact shock energy in the gap between the inner container and the sealing container. The inner container is incorporated with wooden impact shock absorbers being filled so that it is situated in a substantially central portion of an external container. Partitioning cylinders are disposed coaxially in the cylindrical layer filled with wooden impact shock absorbers at an intermediate portion between the outer and the inner containers. Further, a plurality of longitudinally intersecting partitioning disks are disposed each at a predetermined distance in right and left cylindrical wooden impact shock absorbing layers which are in contact with the end face of the inner container. Accordingly, the impact shock energy can be absorbed by the wooden impact shock absorbers efficiently by a plurality of the partitioning disks and the partitioning cylinders. (I.N.)

  12. Pipe line construction for reactor containment buildings

    International Nuclear Information System (INIS)

    Aoki, Masataka; Yoshinaga, Toshiaki

    1978-01-01

    Purpose: To prevent the missile phenomenon caused by broken fragments due to pipe whip phenomenon in a portion of pipe lines connected to a reactor containment from prevailing to other portions. Constitution: Various pipe lines connected to the pressure vessel are disposed at the outside of the containments and they are surrounded with a plurality of protection partition walls respectively independent from each other. This can eliminate the effect of missile phenomena upon pipe rupture from prevailing to the pipe lines and instruments. Furthermore this can afford sufficient spaces for the pipe lines, as well as for earthquake-proof supports. (Horiuchi, T.)

  13. Building Reproducible Science with Singularity Containers

    CERN Multimedia

    CERN. Geneva

    2018-01-01

    Michael Bauer first began working with containers at GSI national lab in Darmstadt, Germany, in 2017 while taking a semester off of school at the University of Michigan. Michael met Greg Kurtzer, project lead of Singularity, during his time at GSI and he began contributing heavily to the Singularity project. At the start of summer 2017, Greg hired Michael to work at the ...

  14. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  15. Radiation doses in buildings containing coal

    International Nuclear Information System (INIS)

    Somlai, J.; Kanyar, B.; Nenyei, A.; Nemeth, Z.; Nemeth, Cs.

    2001-01-01

    Using coal-slag with high concentration of 226 Ra as building material could result excess dose of people living in these dwellings. The gamma dose rate, the radon concentration and the radionuclide concentration of built-in slags were measured in kindergartens, schools and homes of three towns (Ajka, Tatabanya, Varpalota). The absorbed dose rates exceeded significantly the world average (80 nGy/h) and the annual dose reached 3-4 mSv in some cases. The dose coming from radon is significant in the case of slags, which did not originate from power plants but from smaller stoves and furnaces because in these cases the burning temperature is lower, so the radon emanation is higher. The dose in the latter cases could reach 10-20 mSv/year. (author)

  16. Nuclear Capacity Building through Research Reactors

    International Nuclear Information System (INIS)

    2017-01-01

    Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

  17. Failure Mode Estimation of Wolsong Unit 1 Containment Building with respect to Severe Accident Condition

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Choi, In Kil

    2009-01-01

    The containment buildings in a nuclear power plant (NPP) are final barriers against the exposure of harmful radiation materials at severe accident condition. Since the accident at Three Mile Island nuclear plant in 1979, it has become necessary to evaluate the internal pressure capacity of the containment buildings for the assessment of the safety of nuclear power plants. According to this necessity, many researchers including Yonezawa et al. and Hu and Lin analyzed the ultimate capacity of prestressed concrete containments subjected to internal pressure which can be occurred at sever accident condition. Especially in Wolsong nuclear power plant, the Unit 1 containment structures were constructed in the late 1970 to early 1980, so that the end of its service life will be reached in near future. Since that the complete decommission and reconstruction of the NPP may cause a huge expenses, an extension of the service time can be a cost-effective alternative. To extend the service time of NPP, an overall safety evaluation of the containment building under severe accident condition should be performed. In this study, we assessed the pressure capacity of Wolsong Unit 1 containment building under severe accident, and estimated the responses at all of the probable critical areas. Based on those results, we found the significant failure modes of Wolsong Unit 1 containment building with respect to the severe accident condition. On the other hand, for the aged NPP, the degradation of their structural performance must also be explained in the procedure of the internal pressure capacity evaluation. Therefore, in this study, we performed a parametric study on the degradation effects and evaluated the internal pressure capacity of Wolsong Unit 1 containment building with considering aging and degradation effects

  18. Nuclear new build: Financing and project management

    International Nuclear Information System (INIS)

    Keppler, Jan Horst; Cometto, Marco

    2015-01-01

    As there are more nuclear power reactors under construction today than at any time in history, the authors present in a first part an economic and financial analysis comparing the respective exposure of gas-fired and nuclear power generation to electricity price uncertainty and examine the option of leaving the market in the case of a permanent fall in electricity prices. In a second part, the study more specifically addresses the financial risk associated with the development of a new nuclear project, taking into account not only uncertainty about the future evolution of prices but also about the cost of construction and operations, as well as the implications of different ratios of fixed cost to variable cost for bondholders and equity investors. The authors then address the issue of the management of new build projects and their supply chains, as the nuclear industry is undergoing a number of important developments with massive and discontinuous technological changes that are underway as generation II nuclear power plants (NPPs) are substituted by larger and more complex generation III/III+ plants. The loss of skill and human capital as engineers of the nuclear building boom of the 1970's and 1980's retire must also be factored in, as well as the reconfiguration of the global supply chain, which is driven both by new possibilities in data management, externalisation and logistics, and a secular shift of activity from the United States, Japan, Europe and Korea to China, Southeast Asia and the Middle East

  19. Nuclear power plant with a containment

    International Nuclear Information System (INIS)

    Barthelmes, C.P.

    1982-01-01

    In nuclear power plants there is usually a containment incorporating components bearing activity. If in the cladding free hydrogen develops, controlled oxidation must be ensured by means of a recombination device, in order to prevent oxyhydrogen explosions. For this purpose, a permanent thoroughmixing of the gases in the containment is required. This can be achieved by vertical shafts reaching to at least half the height of the containment and provided with heating devices to initiate the gas circulation by the stack effect. These heating devices mainly serve as thermal recombinator. (orig.) [de

  20. Detailed thermal-hydraulic computation into a containment building

    Energy Technology Data Exchange (ETDEWEB)

    Caruso. A.; Flour, I.; Simonin, O. [EDF/LNH, Chatou (France); Cherbonnel, C [EDF/SEPTEN, Villeurbanne (France)

    1995-09-01

    This paper deals with numerical predictions of the influence of water sprays upon stratifications into a containment building using a two-dimensional two-phase flow code. Basic equations and closure assumptions are briefly presented. A test case in a situation involving spray evaporation is first detailed to illustrate the validation step. Then results are presented for a compressible recirculating flow into a containment building with condensation phenomena.

  1. Requirements for a nuclear new build. RWE

    International Nuclear Information System (INIS)

    Mantz, Joern Erik

    2011-01-01

    The investment decisions for large projects need a thoroughly evaluation of the political and economic boundary conditions. The authors discusses the respective questions of nuclear new builds for power utilities like RWE. The public acceptance and the political decisions are difficult to influence by the utility, thus the economic aspects are in the center of consideration. For the economic assessment of a nuclear power plant operational and fuel costs are minor factors, the largest part are the capital costs. Low sunk cost case and vendor selection mechanisms allow the utilities to reach short construction times and thus enhancement of the profit.

  2. Piercing of the containment shell of a reactor building in case of airplane crash

    International Nuclear Information System (INIS)

    Herzog, M.

    1978-01-01

    The author presents a simple calculation model for a realistic check of the piercing safety of containments of reactor buildings in case of airplane crash. Its application is illustrated by a numerical example (Starfighter crash on the Unterweser nuclear power plant). (orig.) [de

  3. Predicting the Lifetimes of Nuclear Waste Containers

    Science.gov (United States)

    King, Fraser

    2014-03-01

    As for many aspects of the disposal of nuclear waste, the greatest challenge we have in the study of container materials is the prediction of the long-term performance over periods of tens to hundreds of thousands of years. Various methods have been used for predicting the lifetime of containers for the disposal of high-level waste or spent fuel in deep geological repositories. Both mechanical and corrosion-related failure mechanisms need to be considered, although until recently the interactions of mechanical and corrosion degradation modes have not been considered in detail. Failure from mechanical degradation modes has tended to be treated through suitable container design. In comparison, the inevitable loss of container integrity due to corrosion has been treated by developing specific corrosion models. The most important aspect, however, is to be able to justify the long-term predictions by demonstrating a mechanistic understanding of the various degradation modes.

  4. Containment vessel construction for nuclear power reactors

    International Nuclear Information System (INIS)

    Sulzer, H.D.; Coletti, J.L.

    1975-01-01

    A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

  5. Building Public Confidence in Nuclear Activities

    International Nuclear Information System (INIS)

    Isaacs, T

    2002-01-01

    Achieving public acceptance has become a central issue in discussions regarding the future of nuclear power and associated nuclear activities. Effective public communication and public participation are often put forward as the key building blocks in garnering public acceptance. A recent international workshop in Finland provided insights into other features that might also be important to building and sustaining public confidence in nuclear activities. The workshop was held in Finland in close cooperation with Finnish stakeholders. This was most appropriate because of the recent successes in achieving positive decisions at the municipal, governmental, and Parliamentary levels, allowing the Finnish high-level radioactive waste repository program to proceed, including the identification and approval of a proposed candidate repository site. Much of the workshop discussion appropriately focused on the roles of public participation and public communications in building public confidence. It was clear that well constructed and implemented programs of public involvement and communication and a sense of fairness were essential in building the extent of public confidence needed to allow the repository program in Finland to proceed. It was also clear that there were a number of other elements beyond public involvement that contributed substantially to the success in Finland to date. And, in fact, it appeared that these other factors were also necessary to achieving the Finnish public acceptance. In other words, successful public participation and communication were necessary but not sufficient. What else was important? Culture, politics, and history vary from country to country, providing differing contexts for establishing and maintaining public confidence. What works in one country will not necessarily be effective in another. Nonetheless, there appear to be certain elements that might be common to programs that are successful in sustaining public confidence and some of

  6. Building Public Confidence in Nuclear Activities

    International Nuclear Information System (INIS)

    Isaacs, T

    2002-01-01

    Achieving public acceptance has become a central issue in discussions regarding the future of nuclear power and associated nuclear activities. Effective public communication and public participation are often put forward as the key building blocks in garnering public acceptance. A recent international workshop in Finland provided insights into other features that might also be important to building and sustaining public confidence in nuclear activities. The workshop was held in Finland in close cooperation with Finnish stakeholders. This was most appropriate because of the recent successes in achieving positive decisions at the municipal, governmental, and Parliamentary levels, allowing the Finnish high-level radioactive waste repository program to proceed, including the identification and approval of a proposed candidate repository site Much of the workshop discussion appropriately focused on the roles of public participation and public communications in building public confidence. It was clear that well constructed and implemented programs of public involvement and communication and a sense of fairness were essential in building the extent of public confidence needed to allow the repository program in Finland to proceed. It was also clear that there were a number of other elements beyond public involvement that contributed substantially to the success in Finland to date. And, in fact, it appeared that these other factors were also necessary to achieving the Finnish public acceptance. In other words, successful public participation and communication were necessary but not sufficient. What else was important? Culture, politics, and history vary from country to country, providing differing contexts for establishing and maintaining public confidence. What works in one country will not necessarily be effective in another. Nonetheless, there appear to be certain elements that might be common to programs that are successful in sustaining public confidence, and some of

  7. Containment vessel for a nuclear reactor

    International Nuclear Information System (INIS)

    Yamanari, Sh.; Horiuchi, T.; Sugisaki, T.; Tominaga, K.

    1985-01-01

    A containment vessel for a nuclear reactor having a dry well for mounting therein a pressure vessel for containing the nuclear reactor, a pressure suppressing chamber having a pool of coolant therein, and a vent pipe device for releasing therethrough into the pool of coolant within the pressure suppressing chamber steam which will be produced as a result of the occurrence of an accident and escape into the dry well. The vent pipe device includes a plurality of vent pipe members inserted in the pool of coolant within the pressure suppressing chamber and each having at least one exhaust port opening in the coolant. The vent pipe members are divided into a plurality of groups in such a manner that the vent pipe members of different groups differ from one another in the length of submerged portions of the vent pipe members interposed between the liquid of the coolant within the pressure suppressing chamber and the exhaust ports of the vent pipe members

  8. Hydraulic behaviour of a representative structural volume for containment buildings

    International Nuclear Information System (INIS)

    Jason, Ludovic; Pijaudier-Cabot, Gilles; Ghavamian, Shahrokh; Huerta, Antonio

    2007-01-01

    For particular structures like containment buildings of nuclear power plants, the study of the hydraulic behaviour is of great concern. These structures are indeed the third barrier used to protect the environment in case of accidents. The evolution of the leaking rate through the porous medium is closely related to the changes in the permeability during the ageing process of the structure. It is thus essential to know the relation between concrete degradation and the transfer property when the consequences of a mechanical loading on the hydraulic behaviour have to be evaluated. A chained approach is designed for this purpose. The mechanical behaviour is described by an elastic plastic damage formulation, where damage is responsible for the softening evolution while plasticity accounts for the development of irreversible strains. The drying process is evaluated according to a non-linear equation of diffusion. From the knowledge of the damage and the degree of saturation, a relation is proposed to calculate the permeability of concrete. Finally, the non-homogeneous distribution of the hydraulic conductivity is included in the hydraulic problem which is in fact the association of the mass balance equation for gas phase and Darcy law. From this methodology, it is shown how an indicator for the hydraulic flows can be deduced

  9. Nanoporous Glasses for Nuclear Waste Containment

    OpenAIRE

    Woignier, Thierry; Primera, Juan; Reynes, Jerôme

    2016-01-01

    Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical dura...

  10. Nuclear power plants and house-building

    International Nuclear Information System (INIS)

    1982-04-01

    In this report, it is studied where and under what conditions houses can be built in the neighbourhood of nuclear power plants. Building norms and location distances are investigated. First, the Dutch and foreign norms with respect to population densities are listed. Only industrial, densely populated neighbour countries are considered. Next, it is calculated what consequences for housing a nuclear station may have within a radius of 20 km. Using these calculations it is studied whether the existing Dutch requirements are satisfied. Because it is expected that the norms are likely to be tightened up under the pressure of the nuclear controversy, the existing Dutch situation is also compared with the tighter foreign norms. Finally, the results of the study are summarized and some conclusions are drawn. (Auth.)

  11. Double containment shell for nuclear power plants

    International Nuclear Information System (INIS)

    Sykora, D.

    1977-01-01

    A double containment shell is proposed for nuclear power plants, especially those equipped with pressurized water reactors. The shell offers increased environmental protection from primary circuit accidents. The inner shell is built of steel or concrete while the outer shell is always built of concrete. The space between the two shells is filled with water and is provided with several manholes and with stiffeners designed for compensation for load due to the water hydrostatic pressure. Water serves the airtight separation of the containment shell inside from the environment and the absorption of heat released in a primary circuit accident. In case the inner shell is made of concrete, it is provided with heat-removal tubes in-built in its walls ensuring rapid heat transfer from the inside of the containment to the water in the interwall space. (Z.M.)

  12. Numerical analyses of an aircraft crash on containment building

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Jae Min; Kim, Seung Hyun; Chang, Yoon Suk [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    The containment building is responsible to isolate and protect internal devices against external conditions like earthquake, hurricane and impact loading. It has also to protect leakage of radioactivity, like LOCA (Loss Of Coolant Accident), when severe accidents occurred. Meanwhile, social awareness such as terrorism has been increased globally after international aircraft crashes at World Trade Center and Pentagon. In this paper, FE (Finite Element) analyses according to variation of crash locations and speeds were carried out to examine the aircraft crash impact on a domestic containment building. In this paper, numerical analyses of aircraft crash on NPP's containment building were performed taking into account different locations and aircraft speeds. (1) Amounts of concrete failure were dependent on the crash locations and the connector was the most delicate location comparing to the dome and wall part. (2) Maximum stress values generated at the liner plate and rebars did not exceed their UTS values.

  13. Numerical analyses of an aircraft crash on containment building

    International Nuclear Information System (INIS)

    Sim, Jae Min; Kim, Seung Hyun; Chang, Yoon Suk

    2016-01-01

    The containment building is responsible to isolate and protect internal devices against external conditions like earthquake, hurricane and impact loading. It has also to protect leakage of radioactivity, like LOCA (Loss Of Coolant Accident), when severe accidents occurred. Meanwhile, social awareness such as terrorism has been increased globally after international aircraft crashes at World Trade Center and Pentagon. In this paper, FE (Finite Element) analyses according to variation of crash locations and speeds were carried out to examine the aircraft crash impact on a domestic containment building. In this paper, numerical analyses of aircraft crash on NPP's containment building were performed taking into account different locations and aircraft speeds. (1) Amounts of concrete failure were dependent on the crash locations and the connector was the most delicate location comparing to the dome and wall part. (2) Maximum stress values generated at the liner plate and rebars did not exceed their UTS values

  14. Building a medical system for nuclear facilities

    International Nuclear Information System (INIS)

    Maeda, Mitsuya

    2016-01-01

    To build a medical system for nuclear facilities, I explained what kinds of actions were performed with the TEPCO Fukushima Daiichi Nuclear Power Plant Accident and what kinds of actions are going to be performed in the future. We examined the health and medical care of the emergency workers in nuclear facilities including TEPCO Fukushima Daiichi Nuclear Power Plant from 2014 to 2015 in the Ministry of Health, Labour and Welfare (MHLW). We carried out a detailed hearing from stakeholders of electric companies and medical institutions about the medical system in nuclear facilities carrying out urgent activities. It has been said that the electric company is responsible to maintain the medical system for affected workers in nuclear facilities. However, TEPCO could not find the medical staff, such as doctors, by their own effort at the TEPCO Fukushima Daiichi Nuclear Power Plant Accident. The network of doctors familiar with emergency medical care support dispatched the medical staff after July of 2011. The stakeholders indicated that the following six tasks must be resolved: (1) the fact that no electric company performs the action of bringing up medical staff who can be dispatched into nuclear facilities in emergencies in 2015; (2) bringing up personnel in charge of radiation management and logistics other than the medical staff, such as doctors; (3) cooperation with the community medicine system given the light and shade by nuclear facilities; (4) performing training for the many concurrent wounded based on the scenario of a severe accident; (5) indicating both the condition of the contract and the guarantee of status that is appropriate for dispatched medical staffs; and (6) clarifying the organization of the network of stakeholders. The stakeholders showed the future directionality as follows: (1) To recruit the medical staff expected to be dispatched into nuclear facilities, (2) to carry out the discussion and conveyance training to strengthen cooperation with

  15. Civil engineering in nuclear power stations: design of the turbine building and nuclear auxiliary building

    International Nuclear Information System (INIS)

    Lacroix, R.

    1985-01-01

    After enumerating the specific features of civil engineering in nuclear power stations. One goes on to examine the principal deliberations undertaken with the aim of optimising projects for transition from the P4 to P'4 and then N4 generations of nuclear power stations. The courses of action decided with respect to the design of the machine room and auxiliary equipment building are described [fr

  16. Standardized analyses of nuclear shipping containers

    International Nuclear Information System (INIS)

    Parks, C.V.; Hermann, O.W.; Petrie, L.M.; Hoffman, T.J.; Tang, J.S.; Landers, N.F.; Turner, W.D.

    1983-01-01

    This paper describes improved capabilities for analyses of nuclear fuel shipping containers within SCALE -- a modular code system for Standardized Computer Analyses for Licensing Evaluation. Criticality analysis improvements include the new KENO V, a code which contains an enhanced geometry package and a new control module which uses KENO V and allows a criticality search on optimum pitch (maximum k-effective) to be performed. The SAS2 sequence is a new shielding analysis module which couples fuel burnup, source term generation, and radial cask shielding. The SAS5 shielding sequence allows a multidimensional Monte Carlo analysis of a shipping cask with code generated biasing of the particle histories. The thermal analysis sequence (HTAS1) provides an easy-to-use tool for evaluating a shipping cask response to the accident capability of the SCALE system to provide the cask designer or evaluator with a computational system that provides the automated procedures and easy-to-understand input that leads to standarization

  17. Ageing degradation in the Gentilly-1 concrete containment building

    International Nuclear Information System (INIS)

    Jaffer, S.; Pentecost, S.; Angell, P.; Shenton, B.

    2015-01-01

    Concrete containment buildings (CCBs) are designed for a service life up to 40 years, but nuclear power plant (NPP) refurbishment can extend service life beyond 60 years. Only limited testing can be conducted on an in-service CCB. The Gentilly-1 (G-1) NPP is in a safe, sustainable shutdown state and the G-1 CCB was available for testing to determine age-related degradation that may be relevant to operating CCBs. Visual observation of the G-1 CCB helped to identify various signs of degradation. However, field testing, via concrete removal, was performed to: (i) examine reinforcing bars and concrete to determine their condition and in-situ stresses and (ii) examine condition of post-tensioned (P-T) wires. The concrete was also subjected to laboratory tests to evaluate its physical, mechanical and chemical properties such as compressive strength, carbonation depth, chloride content and presence of internal degradation. The degradation mechanisms that were clearly visible include macro- and micro-cracking, efflorescence, and weathering. The reinforcing bars in the perimeter wall and dome exposed during the program showed no evidence of active corrosion. Corrosion products were observed on the surfaces of most exposed P-T wires in the perimeter wall, but none were present on P-T wires exposed in the dome. Laboratory testing on the concrete cores extracted from the CCB revealed compressive strength in excess of the design requirements, low carbonation depths (< 10 mm) and no appreciable chlorides. Micro-cracking was observed in the samples recovered from the wall and dome. To date, the observed micro-cracking has had no apparent visible affect on the performance of the CCB concrete. (authors)

  18. Nanoporous Glasses for Nuclear Waste Containment

    Directory of Open Access Journals (Sweden)

    Thierry Woignier

    2016-01-01

    Full Text Available Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical durability, silica glass optimizes the properties of a suitable host matrix. According to an easy sintering stage, nanoporous glasses such as xerogels, aerogels, and composite gels are alternative ways to synthesize silica glass at relatively low temperatures (≈1,000–1,200°C. Nuclear wastes exist as aqueous salt solutions and we propose using the open pore structure of the nanoporous glass to enable migration of the solution throughout the solid volume. The loaded material is then sintered, thereby trapping the radioactive chemical species. The structure of the sintered materials (glass ceramics is that of nanocomposites: actinide phases (~100 nm embedded in a vitreous silica matrix. Our results showed a large improvement in the chemical durability of glass ceramic over conventional nuclear glass.

  19. Analysis of flammability in the attached buildings to containment under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, J.C. de la, E-mail: juan-carlos.de-la-rosa-blul@ec.europa.eu [European Commission Joint Research Centre (Netherlands); Fornós, Joan, E-mail: jfornosh@anacnv.com [Asociación Nuclear Ascó-Vandellós (Spain)

    2016-11-15

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H{sub 2} and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  20. Analysis of flammability in the attached buildings to containment under severe accident conditions

    International Nuclear Information System (INIS)

    Rosa, J.C. de la; Fornós, Joan

    2016-01-01

    Highlights: • Analysis of flammability conditions in buildings outside containment. • Stepwise approach easily applicable for any kind of containment and attached buildings layout. • Detailed application for real plant conditions has been included. - Abstract: Right after the events unfolded in Fukushima Daiichi, the European Union countries agreed in subjecting Nuclear Power Plants to Stress Tests as developed by WENRA and ENSREG organizations. One of the results as implemented in many European countries derived from such tests consisted of mandatory technical instructions issued by nuclear regulatory bodies on the analysis of potential risk of flammable gases in attached buildings to containment. The current study addresses the key aspects of the analysis of flammable gases leaking to auxiliary buildings attached to Westinghouse large-dry PWR containment for the specific situation where mitigating systems to prevent flammable gases to grow up inside containment are available, and containment integrity is preserved – hence avoiding isolation system failure. It also provides a full practical exercise where lessons learned derived from the current study – hence limited to the imposed boundary conditions – are applied. The leakage of gas from the containment to the support buildings is based on separate calculations using the EPRI-owned Modular Accident Analysis Program, MAAP4.07. The FATE™ code (facility Flow, Aerosol, Thermal, and Explosion) was used to model the transport and distribution of leaked flammable gas (H_2 and CO) in the penetration buildings. FATE models the significant mixing (dilution) which occurs as the released buoyant gas rises and entrains air. Also, FATE accounts for the condensation of steam on room surfaces, an effect which acts to concentrate flammable gas. The results of the analysis show that during a severe accident, flammable conditions are unlikely to occur in compartmentalized buildings such as the one used in the

  1. Catalytic hydrogen recombination for nuclear containments

    International Nuclear Information System (INIS)

    Koroll, G.W.; Lau, D.W.P.; Dewit, W.A.; Graham, W.R.C.

    1994-01-01

    Catalytic recombiners appear to be a credible option for hydrogen mitigation in nuclear containments. The passive operation, versatility and ease of back fitting are appealing for existing stations and new designs. Recently, a generation of wet-proofed catalyst materials have been developed at AECL which are highly specific to H 2 -O 2 , are active at ambient temperatures and are being evaluated for containment applications. Two types of catalytic recombiners were evaluated for hydrogen removal in containments based on the AECL catalyst. The first is a catalytic combustor for application in existing air streams such as provided by fans or ventilation systems. The second is an autocatalytic recombiner which uses the enthalpy of reaction to produce natural convective flow over the catalyst elements. Intermediate-scale results obtained in 6 m 3 and 10 m 3 spherical and cylindrical vessels are given to demonstrate self-starting limits, operating limits, removal capacity, scaling parameters, flow resistance, mixing behaviour in the vicinity of an operating recombiner and sensitivity to poisoning, fouling and radiation. (author). 13 refs., 10 figs

  2. Venting device for nuclear reactor container

    International Nuclear Information System (INIS)

    Yamashita, Masahiro; Ogata, Ken-ichi.

    1994-01-01

    An airtight vessel of a venting device of a nuclear reactor container is connected with a reactor container by way of a communication pipeline. A feed water tank is disposed at a position higher than the liquid surface of scrubbing water in the airtight vessel for supplying scrubbing water to the airtight vessel. In addition, a scrubbing water storage tank is disposed at a position hither than the feed water tank for supplying scrubbing water to the feed water tank. Storage water in the feed water tank is introduced into the airtight vessel by the predetermined opening operation of a valve by the pressure exerted on the liquid surface and the own weight of the storage water. Further, the storage water in the scrubbing water storage tank is led into the feed water tank by the water head pressure. The scrubbing water for keeping the performance of the venting device of the reactor container can be supplied by a highly reliable method without using AC power source or the like as a driving source. (I.N.)

  3. Evaluation of high-pressure containment buildings for LMFBR's

    International Nuclear Information System (INIS)

    Armstrong, G.R.

    1981-01-01

    A study was conducted on the use of High Pressure LMFBR Containment Buildings for 1000 MW(e) LMFBRs. Two principal aspects were investigated: accident consequence mitigation and cost. Two types of hypothetical accidents were analyzed to establish consequence mitigation: melt-through and energetic expulsion. Three Containment Building (CB) design pressures were investigated: 69 kPa (10 psig), 207 kPa (30 psig), and 414 kPa (60 psig). Four types of design structures were analyzed to establish cost: steel, steel with confinement building, reinforced concrete, and prestressed/post-tensioned concrete. Results show that: it is within reason that a high pressure containment for a 1000 MW(e) reactor can be fabricated that will retain its integrity during postulated severe hypothetical accidents, if available measures are taken to reduce or prevent hydrogen production and the cost differential between basic high (414 kPa) and low (69 kPa) pressure containments is $10 x 10 6 or less

  4. Dynamic analysis of reactor containment building using axisymmetric finite element model

    International Nuclear Information System (INIS)

    Thakkar, S.K.; Dubey, R.N.

    1989-01-01

    The structural safety of nuclear reactor building during earthquake is of great importance in view of possibility of radiation hazards. The rational evaluation of forces and displacements in various portions of structure and foundation during strong ground motion is most important for safe performance and economic design of the reactor building. The accuracy of results of dynamic analysis is naturally dependent on the type of mathematical model employed. Three types of mathematical models are employed for dynamic analysis of reactor building beam model axisymmetric finite element model and three dimensional model. In this paper emphasis is laid on axisymmetric model. This model of containment building is considered a reinfinement over conventional beam model of the structure. The nuclear reactor building on a rocky foundation is considered herein. The foundation-structure interaction is relatively less in this condition. The objective of the paper is to highlight the significance of modelling of non-axisymmetric portion of building, such as reactor internals by equivalent axisymmetric body, on the structural response of the building

  5. Hydrogen management in nuclear reactor containment

    International Nuclear Information System (INIS)

    Iyer, Kannan

    2014-01-01

    The talk will present the systematic methodology evolved to assess the hydrogen management in nuclear reactor containment during a severe accident. The focus is on the methodology evolved as the full problem is yet to be solved completely. First, the method to quantify mixing of hydrogen is presented. It is demonstrated that buoyancy modified model is adequate to quantify the process satisfactorily. On noting that the hydrogen levels are higher than the safe limits, effort was directed towards mitigating the concentration. Passive Auto-catalytic Recombiners (PAR) were identified as the potential devices for mitigation. Efforts were then directed to model these and a satisfactory one-step reaction derived from a 12 reaction model was evolved. This model was satisfactory when compared with experimental results with hydrogen concentration below 4%. However, the same when extended to hydrogen concentration of 20%, predicts very high concentration thereby indicating the need for experiments at high hydrogen concentration. (author)

  6. Inspection of Nuclear Power Plant Containment Structures

    Energy Technology Data Exchange (ETDEWEB)

    Graves, H.L.; Naus, D.J.; Norris, W.E.

    1998-12-01

    Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

  7. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    International Nuclear Information System (INIS)

    KLEM, M.J.

    2000-01-01

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the

  8. Building nuclear structures : challenges and achievements

    International Nuclear Information System (INIS)

    Gad, V.M.

    1981-01-01

    Reliability and safety are factors of prime importance in construction of civil engineering structures of nuclear facilities. There cannot be any compromise in the strength and life of the structure. This involves rigorous control of: (1) quality of materials and end products, (2) time taken for construction, (3) cost, and also continuing innovation. India has now accumulated more than three decades of experience in nuclear civil engineering and the civil engineering fraternity of India and particularly of the Department of Atomic Energy is now fully capable of designing and construction of all types of structures involved in the nuclear field. Illustrative examples are given. Dome of the CIRUS reactor was constructed in steel plates, but then there was a switch over to reinforced concrete for containment structures and subsequently to prestressed concrete. The aspects taken into consideration of the design to ensure absolute leak tightness are: (1) earthquake safeguards, (2) concrete surface protection, and (3) minimization of cracking in concrete due to pressure loading and shrinkage. Coordination charts are prepared for monitoring time required for various operations and time and motion studies are employed to cut down on construction time. Close control over the cost is kept through internal and external audit, executing the work departmentally or employing an outside agency as the occasion demands and proper selection of materials. Some of the innovations in materials use and construction techniques are mentioned. (K.M.)

  9. Earthquake proof device for nuclear power plant building

    International Nuclear Information System (INIS)

    Okada, Yasuo.

    1991-01-01

    The structure of the present invention enables three dimensional vibration proof, i.e., in horizontal and vertical directions of a reactor container building. That is, each of the reactor container building and a reactor auxiliary building is adapted as an independent structure. The periphery of the reactor container building is surrounded by the reactor auxiliary building. The reactor auxiliary building is supported against the ground by way of a horizontal vibration proof device. The reactor container building is supported against the ground by way of a three-dimensional vibration proof device that prevents vibrations in both of the horizontal directions, and the vertical directions. The reactor container building is connected to the auxiliary building by way of a vertical vibration proof device. With such a constitution, although the reactor container building is vibration proof against both of the horizontal and the vertical vibrations, the vertical vibration proofness is an extension of inherent vertical vibration period. Accordingly, the head of the building undergoes rocking vibrations. However, since the reactor container building is connected to the reactor auxiliary building, the rocking vibrations are prevented by the reactor auxiliary building. As a result, safety upon occurrence of an earthquakes can be ensured. (I.S.)

  10. Cavity pressure history of contained nuclear explosions

    Energy Technology Data Exchange (ETDEWEB)

    Chapin, C E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    Knowledge of pressure in cavities created by contained nuclear explosions is useful for estimating the possibility of venting radioactive debris to the atmosphere. Measurements of cavity pressure, or temperature, would be helpful in evaluating the correctness of present code predictions of underground explosions. In instrumenting and interpreting such measurements it is necessary to have good theoretical estimates of cavity pressures. In this paper cavity pressure is estimated at the time when cavity growth is complete. Its subsequent decrease due to heat loss from the cavity to the surrounding media is also predicted. The starting pressure (the pressure at the end of cavity growth) is obtained by adiabatic expansion to the final cavity size of the vaporized rock gas sphere created by the explosion. Estimates of cavity size can be obtained by stress propagation computer codes, such as SOC and TENSOR. However, such estimates require considerable time and effort. In this paper, cavity size is estimated using a scheme involving simple hand calculations. The prediction is complicated by uncertainties in the knowledge of silica water system chemistry and a lack of information concerning possible blowoff of wall material during cavity growth. If wall material blows off, it can significantly change the water content in the cavity, compared to the water content in the ambient media. After cavity growth is complete, the pressure will change because of heat loss to the surrounding media. Heat transfer by convection, radiation and conduction is considered, and its effect on the pressure is calculated. Analysis of cavity heat transfer is made difficult by the complex nature of processes which occur at the wall where melting, vaporization and condensation of the gaseous rock can all occur. Furthermore, the melted wall material could be removed by flowing or dripping to the cavity floor. It could also be removed by expansion of the steam contained in the melt (blowoff) and by

  11. Dynamic containment of gaseous effluents in the auxiliary buildings and reinjection of liquid effluents from these buildings back into the reactor building for 900 MWe PWRs under accident condition

    International Nuclear Information System (INIS)

    Demoulin, F.; Collinet, J.; Nguyen, C.

    1987-04-01

    Examination of the lessons to be learned from the accident of the Three Mile Island nuclear power plant on 20 March 1979 led the French Safety Authorities and EDF (Electricite de France) to adopt a series of measures intended to improve the performance of the containment of French PWRs, especially in the event of accident. Among the measures adopted, two of them contribute to the upgrading of the containment of nuclear island buildings, by reducing radioactivity constraints inside these buildings and by limiting radioactive releases into the environment. These are: (1) dynamic containment of auxiliary buildings likely to be contaminated following an accident, (2) reinjection back into the reactor building of liquid effluents arising in the auxiliary buildings. In this paper we shall discuss, for each measure, the approach to the problem and describe the arrangements made to arrive at a satisfactory solution [fr

  12. Investigation of superstructure damping identification for the HDR containment building

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Kot, C.A.; Srinivasan, M.G.

    1985-01-01

    A method for the estimation of first mode structural damping, developed by other investigators, was applied to shaker test data of the HDR containment building. Due to inadequate precision in the experimental phase measurements no valid results could be obtained. Based on modal analysis it was also noted that for systems such as the HDR building, contributions of higher modes are not negligible as was assumed in the original approach. Therefore, the procedure for the determination of superstructure damping using experimental data was extended to include the effects of higher modes. The extended method does not lead to any higher order nonlinear equations than the first mode approximation and was found to be as simple to apply as the original approach

  13. New nuclear build programs status and prospects

    International Nuclear Information System (INIS)

    Tankosic, D.; Sabinov, S.

    2010-01-01

    Presentation topics: Reactor technology update OECD OEMs development; Nuclear power sector development projections - global nuclear capacity projections to 2050 (NEA low , high and phase-out scenarios); new nuclear constructions Worley Parsons support to new nuclear Programs around the world; Drivers towards increased nuclear capacity - increased demand of generating capacities; climate change concerns; Security of supply; Economics; Robust to fuel price increases

  14. Iodine removal in containment filtered venting system during nuclear accident

    International Nuclear Information System (INIS)

    Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

  15. Preliminary CFD Analysis for HVAC System Design of a Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Son, Sung Man; Choi, Choengryul [ELSOLTEC, Yongin (Korea, Republic of); Choo, Jae Ho; Hong, Moonpyo; Kim, Hyungseok [KEPCO Engineering and Construction, Gimcheon (Korea, Republic of)

    2016-10-15

    HVAC (Heating, Ventilation, Air Conditioning) system has been mainly designed based on overall heat balance and averaging concepts, which is simple and useful for designing overall system. However, such a method has the disadvantage that cannot predict the local flow and temperature distributions in a containment building. In this study, a CFD (Computational Fluid Dynamics) preliminary analysis is carried out to obtain detailed flow and temperature distributions in a containment building and to ensure that such information can be obtained via CFD analysis. This approach can be useful for hydrogen analysis in an accident related to hydrogen released into a containment building. In this study, CFD preliminary analysis has been performed to obtain the detailed information of the reactor containment building by using the CFD analysis techniques and to ensure that such information can be obtained via CFD analysis. We confirmed that CFD analysis can offer enough detailed information about flow patterns and temperature field and that CFD technique is a useful tool for HVAC design of nuclear power plants.

  16. Method for building a nuclear power plant and layout thereof

    International Nuclear Information System (INIS)

    Matsuura, Tadashi; Ushiroda, Kouchi; Tajiri, Akinori; Yoshida, Naoto; Takeda, Masakado; Makita, Tasuo; Maezawa, Sumito; Yoshizaki, Masatoshi.

    1997-01-01

    The present invention relates to a technical field of building a nuclear power plant, and its technical issue is to reduce the scale of investment for cranes to be used in building a plurality of nuclear power plant units. In order to resolve the technical issue, in the present invention, a nuclear power plant building method is employed in building a plurality of nuclear plant units in which while an earlier building unit out of the plurality of units which is started to be built earlier is being built, a later building unit that is to be started to be built later than the earlier building unit is started to be built, and in which a common mobile crane is adapted to move between a crane operating area for the earlier building unit and a crane operating area for the later building unit so that the common mobile crane ban be used in building the plurality of units. The present invention is primarily intended for the construction of a nuclear power plant. (author) figs

  17. Nuclear Futures Analysis and Scenario Building

    International Nuclear Information System (INIS)

    Arthur, E.D.; Beller, D.; Canavan, G.H.; Krakowski, R.A.; Peterson, P.; Wagner, R.L.

    1999-01-01

    This LDRD project created and used advanced analysis capabilities to postulate scenarios and identify issues, externalities, and technologies associated with future ''things nuclear''. ''Things nuclear'' include areas pertaining to nuclear weapons, nuclear materials, and nuclear energy, examined in the context of future domestic and international environments. Analysis tools development included adaptation and expansion of energy, environmental, and economics (E3) models to incorporate a robust description of the nuclear fuel cycle (both current and future technology pathways), creation of a beginning proliferation risk model (coupled to the (E3) model), and extension of traditional first strike stability models to conditions expected to exist in the future (smaller force sizes, multipolar engagement environments, inclusion of actual and latent nuclear weapons (capability)). Accomplishments include scenario development for regional and global nuclear energy, the creation of a beginning nuclear architecture designed to improve the proliferation resistance and environmental performance of the nuclear fuel cycle, and numerous results for future nuclear weapons scenarios

  18. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  19. Corrosion process studies in a nuclear waste container

    International Nuclear Information System (INIS)

    Guasp, Ruben A.; Lanzani, Liliana A.; Coronel, Pascual; Bruzzoni, Pablo; Semino, Carlos J.

    1999-01-01

    Latest results on corrosion behavior studies on high activity nuclear waste container are reported. Corrosion evaluation on lead base alloys and modeling to predict carbon steel external container cover generalized corrosion, are the main issues of these studies. (author)

  20. Axisymmetric modeling of prestressing tendons in nuclear containment dome

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Se-Jin [DAEWOO E and C, Institute of Construction Technology, 60 Songjook-dong, Jangan-gu, Suwon, Kyonggi 440-210 (Korea, Republic of)]. E-mail: jsj@dwconst.co.kr; Chung, Chul-Hun [Department of Civil and Environmental Engineering, Dankook University, San 8, Hannam-dong, Youngsan-gu, Seoul 140-714 (Korea, Republic of)

    2005-12-15

    Simple axisymmetric modeling of a nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings such as internal pressure. In this case, the prestressing tendons placed in the containment dome should be axisymmetrically approximated, since most dome tendons are not arranged in an axisymmetric manner. Some procedures are proposed that can realistically implement the actual three-dimensional tendon stiffness and prestressing effect into the axisymmetric model. Prestressing tendons, which are arranged in two or three ways depending on a containment type, are converted into the equivalent layer to consider the stiffness contribution in meridional and hoop directions. In order to reflect the prestressing effect, the equivalent load method and the initial stress method are devised, respectively, and the corresponding loads or stresses are derived in terms of the axisymmetric model. The proposed schemes are verified through some numerical examples comparing the results of the axisymmetric models to those of the actual three-dimensional model. The examples show that the proper level of the prestressing in the hoop direction of the axisymmetric dome plays an important role in tracing the actual behavior induced by the prestressing. Finally, some correction factors are discussed that can further improve the analysis results.

  1. Axisymmetric modeling of prestressing tendons in nuclear containment dome

    International Nuclear Information System (INIS)

    Jeon, Se-Jin; Chung, Chul-Hun

    2005-01-01

    Simple axisymmetric modeling of a nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings such as internal pressure. In this case, the prestressing tendons placed in the containment dome should be axisymmetrically approximated, since most dome tendons are not arranged in an axisymmetric manner. Some procedures are proposed that can realistically implement the actual three-dimensional tendon stiffness and prestressing effect into the axisymmetric model. Prestressing tendons, which are arranged in two or three ways depending on a containment type, are converted into the equivalent layer to consider the stiffness contribution in meridional and hoop directions. In order to reflect the prestressing effect, the equivalent load method and the initial stress method are devised, respectively, and the corresponding loads or stresses are derived in terms of the axisymmetric model. The proposed schemes are verified through some numerical examples comparing the results of the axisymmetric models to those of the actual three-dimensional model. The examples show that the proper level of the prestressing in the hoop direction of the axisymmetric dome plays an important role in tracing the actual behavior induced by the prestressing. Finally, some correction factors are discussed that can further improve the analysis results

  2. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

    Science.gov (United States)

    Zhou, Jianguo; Xu, Yaming; Zhang, Tao

    2016-06-14

    Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

  3. Nuclear waste storage container with metal matrix

    International Nuclear Information System (INIS)

    Sump, K.R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties

  4. Seal welded cast iron nuclear waste container

    International Nuclear Information System (INIS)

    Filippi, A.M.; Sprecace, R.P.

    1987-01-01

    An article of manufacture is described comprising a cast iron container having an opening at one end and a cast iron plug; a first nickel-carbon alloy fusion weldable insert surrounding the opening and metallurgically bonded to the cast iron container at the one end of the container; a second nickel-carbon alloy insert metallurgically bonded to the cast iron plug located within the opening and surrounded by the first insert the inserts being jointed by a fusion bond in the opening without heating the cast iron container to an austenite formation temperature thereby sealing the interior of the container from the exterior ambient outside the opening; the nickel-carbon alloy containing about 2 to 5 w% carbon; and both the nickel-carbon alloy insert and the cast iron container have a microstructure containing a graphite phase

  5. Nuclear waste storage container with metal matrix

    Science.gov (United States)

    Sump, Kenneth R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

  6. Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

    International Nuclear Information System (INIS)

    Park, S.Y.; Ahn, K.I.

    2014-01-01

    Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

  7. Complete analysis of a nuclear building to nuclear safety standards

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, T A

    1975-01-01

    The nuclear standards impose on the designer the necessity of examining the loads, stresses, and strains in a nuclear building even under extreme loading conditions, both due to plant malfunctions and environmental accidents. It is necessary then to generate, combine, and examine a tremendous amount of data; really the lack of symmetry and general complication of the structures and the large number of loading combinations make an automatic analysis quite necessary. A largely automated procedure is presented in view of solving the problem by a series of computer programs linked together. After the seismic analysis has been performed by (SADE CODE) these data together with the data coming from thermal specifications, weight, accident descriptions etc. are fed into a finite element computer code (SAP4) for analysis. They are processed and combined by a computer code (COMBIN) according to the loading conditions (the usual list in Italy is given and briefly discussed), so that for each point (or each selected zone) under each loading condition the applied loads are listed. These data are fed to another computer code (DTP), which determines the amount of reinforcing bars necessary to accommodate the most severe of the loading conditions. The Aci 318/71 and Italian regulation procedures are followed; the characteristics of the program are briefly described and discussed. Some particular problems are discussed, e.g. the thermal stresses due to normal and accident conditions, the inelastic behavior of some frame elements (due to concrete cracking) is considered by means of an 'ad hoc' code. Typical examples are presented and the results are discussed showing a relatively large benefit in considering this inelastic effect.

  8. The spectroscopic study of building composites containing natural sorbents.

    Science.gov (United States)

    Król, M; Mozgawa, W

    2011-08-15

    This work presents the results of FT-IR spectroscopic studies of heavy metal cations (Ag(+), Pb(2+), Zn(2+), Cd(2+) and Cr(3+)) immobilization from aqueous solutions on natural sorbents. The sorption has been conducted on sodium forms of zeolite (clinoptilolite) and clay minerals (mixtures containing mainly montmorillonite and kaolinite) which have been separated from natural Polish deposit. In the next part of the work both sorbents were used to obtain new building composites. It was proven those heavy metal cations' sorption causes changes in IR spectra of the zeolite and clay minerals. These alterations are dependent on the way the cations were sorbed. In the case of zeolite, variations of the bands corresponding to the characteristic ring vibrations have been observed. These rings occur in pseudomolecular complexes 4-4-1 (built of alumino- and silicooxygen tetrahedra) which constitute the secondary building units (SBU) and form spatial framework of the zeolite. The most significant changes have been determined in the region of pseudolattice vibrations (650-700 cm(-1)). In the instance of clay minerals, changes in the spectra occur at two ranges: 1200-800 cm(-1)--the range of the bands assigned to asymmetric Si-O(Si,Al) and bending Al-OH vibrations and 3800-3000 cm(-1)--the range of the bands originating from OH(-) groups stretching vibrations. Next results indicate possibilities of applying the used natural sorbents for the obtainment of new building materials having favourable composition and valuable properties. The zeolite was used for obtaining autoclaved materials with an addition of CaO, and the clay minerals for ceramic sintered materials with an addition of quartz and clinoptilolite were produced. FT-IR studies were also conducted on the obtained materials. Copyright © 2010 Elsevier B.V. All rights reserved.

  9. The theoretical modelling of aerosol behaviour within containment buildings

    International Nuclear Information System (INIS)

    Dunbar, I.H.

    1988-01-01

    The modelling of the deposition of aerosol particles within the containment building plays an important part in determining the effectiveness of the building in reducing releases of activity following accidents. This paper describes attempts to ensure the accuracy of computer codes which model aerosol behaviour, with special reference to the code AEROSIM-M. Code intercomparisons have been used to test the reliability of the coding and the accuracy of the numerical methods. Those codes which assume that the particle size distribution is always lognormal give significantly different results from those which do not make this assumption but instead discretise the range of particle sizes. When the same physical assumptions are made, the predictions of different discrete codes are in reasonable agreement. In comparisons between an earlier version of AEROSIM and sodium fire experiments, the code achieved good agreement on the overall time-scale of deposition. An extensive set of tests of AEROSIM-M against experiments relevant to LWR conditions is underway. (author)

  10. Behavior of cracked concrete nuclear containment vessels during earthquakes

    International Nuclear Information System (INIS)

    Gergely, P.; Stanton, J.F.; White, R.N.

    1975-01-01

    When pressure builds up in a reinforced concrete nuclear containment shell, its cylindrical wall cracks vertically and horizontally at intervals of about five feet. If an earthquake occurs simultaneously with this pressurization, inertia forces are transmitted across the horizontal crack planes. The forces and deformations must be small enough to maintain the integrity of the steel liner. A typical containment shell has a radius of about 65 ft. and a wall thickness of about 4 ft. It is heavily reinforced with vertical, horizontal, and sometimes diagonal bars. A steel shell of about 3 / 8 in. thickness is attached to the concrete with anchors. The seismic shear forces are transmitted across the horizontal cracks by interface shear transfer (combination of shear friction and aggregate interlocking), by dowel action of the bars, and by diagonal bars if they are used. One important question in the design of such vessels is whether the diagonal bars are necessary. In the experimental portion of the current investigation several types of tests were conducted to study the load-slip characteristics of interface shear transfer under high intensity cyclic loading. In some cases external bars provided the clamping action of reinforcement, in more recent tests large diameter embedded bars were used. This presentation summarizes the analytical part of the investigation. A representative load-slip curve has been used in the analyses to assess the intensity of the stresses and deformations, and to study the importance of the variables as an aid in planning future tests

  11. Nuclear Resonance Fluorescence and Isotopic Mapping of Containers

    Science.gov (United States)

    Johnson, Micah S.; McNabb, Dennis P.

    2009-03-01

    National security programs have expressed interest in developing systems to isotopically map shipping containers, fuel assemblies, and waste barrels for various materials including special nuclear material (SNM). Current radiographic systems offer little more than an ambiguous density silhouette of a container's contents. In this paper we will present a system being developed at LLNL to isotopically map containers using the nuclear resonance fluorescence (NRF) method. Recent experimental measurements on NRF strengths in SNM are discussed.

  12. Innovative Nuclear Power Plant Building Arrangement in Consideration of Decommissioning

    OpenAIRE

    Won-Jun Choi; Myung-Sub Roh; Chang-Lak Kim

    2017-01-01

    A new concept termed the Innovative Nuclear Power Plant Building Arrangement (INBA) strategy is a new nuclear power plant building arrangement method which encompasses upfront consideration of more efficient decommissioning. Although existing decommissioning strategies such as immediate dismantling and differed dismantling has the advantage of either early site restoration or radioactive decommissioning waste reduction, the INBA strategy has the advantages of both strategies. In this research...

  13. Development of degradation D/B system for the containment building of NPP

    International Nuclear Information System (INIS)

    Cho, M. S.; Song, Y. C.; Yim, J. H.; Kim, D. K.; Lee, J. S.

    2001-01-01

    The Degradation D/B System is developed for digitalizing the history of the Containment building of nuclear power plant. It have 6 D/B which are consist of General, Design drawing, Material, Construction, ISI·SIT·ILRT D/B. For efficient operation of the system, utilities are also developed such as the aging and repair data management program for concrete and steel structures, the data search engine with various options helping users find what they want, and the data exchange program restoring and updating input data

  14. Analysis of impact of large commercial aircraft on a prestressed containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyoungsoo, E-mail: kylee@pvamu.edu [Center for Energy and Environmental Sustainability, Prairie View A and M University, Prairie view, TX, 77446 (United States); Han, Sang Eul, E-mail: hsang@inha.ac.kr [Department of Architectural Engineering, School of Architecture, Inha University, 253 Yonghyundong Nam-gu, Incheon, 402-751 (Korea, Republic of); Hong, Jung-Wuk, E-mail: j.hong@kaist.ac.kr [Department of Civil and Environmental Engineering, KAIST, 373-1 Guseon-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

    2013-12-15

    Highlights: • Aircraft impact analyses are performed using the missile–target interaction method. • A large commercial B747 aircraft is considered with erosion effect. • The rigid wall impact test shows the validity of the developed aircraft model. • The parametric studies on the fictitious containment building are performed. • The plastic failure of the target is governed by the impulse of aircraft at the first momentum peak. - Abstract: In this paper, the results of nonlinear dynamic analyses of a concrete containment building under extreme loads are presented. The impact of a large commercial B747 airliner is investigated as the extreme load, and a rigid wall impact test is performed using commercial nonlinear finite element codes. The impact forces exerted by the aircraft are verified compared with the time-dependent impact force provided by OECD/NEA (2002), which was calculated based on the so-called Riera method. The rigid wall impact analysis shows that the finite element model of a B747 is appropriate for the purpose of the aircraft crash analysis exposed to the external hazard of “Beyond Design-Basis Events” defined by U.S. Nuclear Regulatory Commission. Finally, the applicability of this methodology is further studied and verified by conducting parametric studies on the critical infrastructures of nuclear power plant containment structures.

  15. Analysis of impact of large commercial aircraft on a prestressed containment building

    International Nuclear Information System (INIS)

    Lee, Kyoungsoo; Han, Sang Eul; Hong, Jung-Wuk

    2013-01-01

    Highlights: • Aircraft impact analyses are performed using the missile–target interaction method. • A large commercial B747 aircraft is considered with erosion effect. • The rigid wall impact test shows the validity of the developed aircraft model. • The parametric studies on the fictitious containment building are performed. • The plastic failure of the target is governed by the impulse of aircraft at the first momentum peak. - Abstract: In this paper, the results of nonlinear dynamic analyses of a concrete containment building under extreme loads are presented. The impact of a large commercial B747 airliner is investigated as the extreme load, and a rigid wall impact test is performed using commercial nonlinear finite element codes. The impact forces exerted by the aircraft are verified compared with the time-dependent impact force provided by OECD/NEA (2002), which was calculated based on the so-called Riera method. The rigid wall impact analysis shows that the finite element model of a B747 is appropriate for the purpose of the aircraft crash analysis exposed to the external hazard of “Beyond Design-Basis Events” defined by U.S. Nuclear Regulatory Commission. Finally, the applicability of this methodology is further studied and verified by conducting parametric studies on the critical infrastructures of nuclear power plant containment structures

  16. Nuclear reactor melt-retention structure to mitigate direct containment heating

    International Nuclear Information System (INIS)

    Tutu, N.K.; Ginsberg, T.; Klages, J.R.

    1991-01-01

    This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

  17. Why do we build nuclear power stations

    International Nuclear Information System (INIS)

    Keppler, E.

    1984-01-01

    The author discusses some aspects of interest in the context of opposition to nuclear power in Switzerland. The part played and to be played by nuclear power in Switzerland is discussed, criticisms and objections are countered, the implications of power generation without further nuclear contribution are examined, and requests in certain quarters for a tax on power generation except from alternative sources are rejected. (P.G.R.)

  18. Vent control device for nuclear reactor container

    International Nuclear Information System (INIS)

    Kubota, Ryuji.

    1989-01-01

    The present invention concerns automatic prevention of abnormal over-pressure and hydrogen gas flashing in a BWR type reactor container. That is, (1) if the pressure in the container is abnormally increased, the gas in the pressure suppression chamber is released to reduce the pressure thereby preventing over-pressure damage to the container. (2) Then, if exhaust gases are burnt to cause flashing explosion danger for the gases in the reactor container, the gas release is interrupted. The foregoing two functioins are automatically conducted in this device. Specifically, when the pressure in the reactor container reaches a predetermined allowable limit, a remote control operation valve is opened by automatic control means to release the gas in the vessel. Since the gas flow rate at the start of the release exceeds flame propagation velocity, there is no worry for flashing explosion. Further, if the pipeway flow velocity near the atmospheric release is reduced to less than the flame propagation velocity of the hydrogen gas, the opened valve is automatically closed. Accordingly, propagation of hydrogen gas flame into the container thus causing explosion can surely be prevented. (K.M.)

  19. Building infrastructure for new nuclear power programmes

    International Nuclear Information System (INIS)

    Starz, A.; Aoki, M.

    2010-01-01

    In recent years, more than sixty countries have indicated that they are considering or launching nuclear power programmes. It has been more than a decade since a country commissioned its first nuclear power plant. In meantime, the global nuclear community has faced greater concerns about safety, security and non-proliferation, resulting in increased international obligations and a greater expectation for transparency and openness regarding nuclear power programmes. Many of these 'nuclear newcomers' are turning to International Atomic Energy Agency (IAEA) to understand the implications of the nuclear power option and to receive advice about how to proceed with implementing a national programme. In response to growing demand for assistance, the IAEA developed a comprehensive, phased approach to establishing the infrastructure necessary to support a national nuclear power programme. This 'Milestones' approach is described in Nuclear Energy Series Guide NG-G-3.1 'Milestones in the Development of a National Infrastructure for Nuclear Power' (2007). From establishing the national position and legal framework to nuclear safety, security and safeguards, the Milestones covers 19 issues that need to be addressed. This approach also places special emphasis on the need for involvement of the Government, utility, industry, academic, and other stakeholders in a national decision-making process. The IAEA is also helping 'newcomers' to better understand its Safety Standards, which were written from the perspective of operating nuclear power programmes. A new safety guide is in development which provides a Road-map to the safety standards and identifies the standards that are relevant for each phase consistent with the Milestones. Several countries in the Europe region are working with the IAEA to understand the issues associated with a nuclear power programme in preparation for making a knowledgeable commitment. The starting points and approaches vary widely: some are European

  20. PREST, Pressure Temperature Transients, I Inhalation in Containment Building from LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Gaggero, G [CETIS, EURATOM C.C.R., 21020 - Ispra - Varese (Italy); Gerini, P M [CISE, Segrate, Milano (Italy); Leoni, G [AGIP Nucleare, San Donato Milanese - Milano (Italy); Van Erp, J B [EURATOM C.C.R., 21020 - Ispra - Varese (Italy)

    1969-06-01

    1 - Nature of physical problem solved: The programme is intended for the determination of pressure and temperature transient inside the containment building, following a loss-of-coolant accident due to a rupture in the primary cooling system of a nuclear power plant having water as the primary coolant. The model includes the calculation of the radiation doses incurred to the thyroid due to inhalation of radioactive iodine released outside the containment building. 2 - Method of solution: The energy equation is solved at each time step by using the Newton method. In order to determine the heat exchange with structures inside the containment building as well as with the outside atmosphere, the structures are treated in slab geometry. The resulting Fourier equations for heat conduction are solved numerically by using an implicit form to avoid stability problems. 3 - Restrictions on the complexity of the problem: max. number of internal slabs - 6; max. number of external slabs - 4; max. number of meshes in each slab - 100.

  1. Influence of temperature on strain monitoring of degradation in concrete containment buildings

    International Nuclear Information System (INIS)

    Ding, Y.; Jaffer, S.; Angell, P.

    2015-01-01

    Concrete containment buildings (CCBs) are important safety structures in a nuclear power plant (NPP). The CCBs can be made of reinforced and post-tensioned (P-T) concrete. Post-tensioning concrete induces compressive stresses, which have to be overcome for the concrete to crack under tensile loads. However, post-tensioned CCBs may undergo pre-stressing losses as they age, which could affect their performance under accident conditions. CANDU 6 reactor buildings contain grouted post-tensioned tendons as the primary reinforcement. The grouting of the tendons makes direct monitoring of pre-stressing losses via lift-off testing impossible. Therefore, instruments have been installed on an existing reactor building to measure and monitor strains and stresses in the concrete and the deformation of the concrete structure to detect aging degradation and indirectly evaluate the pre-stressing losses. However, the instrumentation readings are affected by temporary volume changes in the concrete caused by the influence of environmental factors, particularly temperature, on concrete. In this work, the focus is on developing an understanding of the effect of temperature on the interpretation of instrumentation data from a reactor building. Vibrating Wire Strain Gauge (VWSG) data has been analysed. The influence of concrete coefficient of thermal expansion and temperature distribution within the reactor building walls, on VWSG data, is discussed based on the analysis of the available instrumentation data and available numerical simulation results. The present study demonstrates that temperature distribution within the containment concrete has a significant impact on the VWSG measurements and the coefficient of thermal expansion of concrete is an important factor in the correction of VWSG data for thermal strain. It is recommended that VWSG data obtained over small temperature variations be considered for interpretation to assess pre-stressing losses. (authors)

  2. Seal for an object containing nuclear fuel

    International Nuclear Information System (INIS)

    Scheuerpflug, W.; Nentwich, D.

    1977-01-01

    This seal which cannot be counterfeited, specially for sealing nuclear objects, e.g. fuel rods, not only makes any damage which has taken place obvious, but makes identification according to a key possible. For this purpose a minimum number of 'particles' or small bodies, which are identical but of different permeability, are fixed inside a short tube during 'loading' of the seal in a certain or an accidental sequence. The sequence of the spheres, which represents a key, can only be determined by special electromagnetic measuring equipment. On first opening the seal, this key sequence is irrevocably destroyed. (HP) [de

  3. Nuclear containment systems and in-service inspection status of Korea nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

    2007-07-01

    20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.

  4. Fuel containing vessel for transporting nuclear fuel

    International Nuclear Information System (INIS)

    Yoshizawa, Hiroyasu; Shimizu, Fukuzo; Tanaka, Nobuyuki.

    1996-01-01

    A shock absorbing mechanism is disposed on an inner bottom of a vessel main body. The shock absorbing mechanism comprises a shock absorbing member disposed on the upper surface of a bottom wall, an annular metal plate disposed on the upper surface of the shock absorbing member and an annular spacer disposed on the upper surface of the metal plate. The shock absorbing member is made of a material such as of wood, lead, metal honeycomb or a metal mesh, which plastically deforms when applied with load higher than a predetermined level, and is formed in a square block-like form covering the upper surface of the bottom wall. The spacer is made of a thin soft material such as tetrafluoroethylene, and is formed in such a shape as capable of preventing direct contact of the lower end of the cylindrical member in a lower tie plate of nuclear fuels with the metal portion. This can ensure integrity of nuclear fuels even when they fall from a high place upon an assumed dropping accident. (I.N.)

  5. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  6. Influence of heat transfer on walls due to aerosol decomposition rate in the containment building of nuclear power plants during heavy incidents; Einfluss des Waermeuebergangs an Waenden auf die Aerosolabbaurate im Sicherheitsbehaelter von Kernkraftwerken bei schweren Stoerfaellen

    Energy Technology Data Exchange (ETDEWEB)

    Braun, T.

    2004-07-01

    Today, German nuclear power plants are leading in safety standards worldwide. Increasing potentials arise continuously along with improvements in technology. One of these potentials is the best-estimate simulation of fission product transport in case of a severe accident. A main part of the fission products is allocated on aerosols. Therefore, the aerosol behavior before containment leakage is important for the radioactive source term to the environment. Having a good knowledge about the main aerosol phenomena, it is possible to simulate them numerically. This enables to develop and test safety measures to limit damages before accidents occur. Within this study, the main aerosol phenomena have been ascertained and accordingly classified into formation, transport and reduction. On this basis, simulations of one- and multi-component aerosol experiments of the KAEVER series have been performed with the COCOSYS code. Due to an overprediction of the computed volume condensation rate, the results showed an overestimation of the reduction rate of insoluble aerosols. The reason was found to be the underestimation of the wall condensation rate. Based on an additional plain thermal hydraulic multi compartment experiment, these uncertainties in the wall heat transfer correlations were investigated in detail. The results show a strong dependency between the wall condensation rate and the convective heat transfer, resp. the characteristic length. In case of mainly forced convection, correct values for the characteristic length led to an underestimation of the calculated heat transfer coefficients. The analysis of the heat transfer models show an inconsistency in the coupling of free and forced convection. Therefore, an improved and consistent convection model has been developed and implemented. Both models have been tested on different experiments. Although the new model shows only minor improvements, it could be proven that the influence for forced convection is significant

  7. Full autonomous monitoring tools inside nuclear reactor building

    International Nuclear Information System (INIS)

    Giraud, A.; Veau, J.

    2009-01-01

    In this paper, we define, design and test a radiation tolerant autonomous monitoring tool for nuclear embedded applications. The goal of the instrumentation system was to record the values of some parameters such as dose, temperature or vibrations appearing inside the containment building of nuclear power plants. The knowledge of these parameters will be a good help for predictive maintenance of the power plant components. For the design of the monitoring tool, we rely on commercial-off-the-shelf (COTS) low power electronic components to use battery-supplied power. A large amount of components starting from discrete transistors or logic units to memories and micro-controllers was associated to define and design a prototype. We then confirm the environment conditions tolerance estimated to up to 2 kGy of total dose and 80 C for temperature by on-line irradiation experiments for individual components and functions and prototypes. Two different sets of about 60 systems were realised as industrial products and then installed in EDF facilities. They were exploited during NPPs operating times of about eighteen months. This large scale experiment was concluded with the inspection of some withdrawn systems. No major degradations were observed showing the efficiency of the hardening method allowing regular use of COTS

  8. Halon 1301 protection system for nuclear containments

    International Nuclear Information System (INIS)

    McHale, E.T.

    1981-01-01

    Halon 1301 can provide protection against any combustion hazard that hydrogen gas might present in an LWR containment following a loss-of-coolant accident. A development program was conducted, comprising analytical study, laboratory experiments and large-scale testing, to define the requirements for a Halon 1301 system and to examine certain operational problems that were hypothesized. Some results of the study are presented in this paper

  9. Managing key capabilities: A challenge for nuclear plant building companies

    International Nuclear Information System (INIS)

    Pascal Corbel

    2003-01-01

    The nuclear plant building industry faces a paradoxical situation. The use of nuclear reactors to produce energy for civil purposes is both a promising technology, with potentially huge outlets, and a technology facing declining demand. One of the key problems is then: how to maintain the capabilities necessary to benefit from the potential recovery? The resource-based view of strategic management has shown the importance of different types of resources and capabilities in gaining a sustainable competitive advantage. Successful incumbents in the market of nuclear station building have built those kinds of distinctive capabilities that give them a competitive advantage over potential new entrants. But we show that, without a permanent activity in plant building, preserving those capabilities necessitates specific strategic action. We firstly develop the argument that the nuclear plant building industry is in a paradoxical situation in terms of demand and technical performance trends. Secondly, we try to identify the key capabilities of the incumbents. We show that companies in that field use mainly three types of distinctive capabilities: pure technical and scientific knowledge in direct relation to the use of nuclear as an energy generator, competences in risk management and competences in large project management, including financing. Thirdly, we show that although some of those capabilities are used through other nuclear-related activities such as plant maintenance or fuel supply, some of them necessitate taking strategic actions in order to be preserved. We argue that this should be a priority of nuclear equipment company managers in the next few years. (author)

  10. Severe accidents and nuclear containment integrity (SANCY). SANCY summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I. [VTT Processes, Espoo (Finland)

    2004-07-01

    SANCY project investigates physical phenomena related to severe nuclear accidents with importance to Finnish nuclear power plants. Currently the major topics are the ex-vessel coolability issues, long-term severe accident management and containment leak tightness and adoption and development of new calculation tools considering also the needs of the future Olkiluoto 3 plant. SANCY employs both experimental and analytical methods. (orig.)

  11. International collaboration for nuclear competence building

    International Nuclear Information System (INIS)

    Haapalehto, T.; Storey, P.

    2004-01-01

    The life cycle of the nuclear industry is no different to that of any other industry, indeed to most forms of human activity: birth, growth, maturity, decline, rebirth and renewal or death. As a result of the twin facets of long time scales and essential technical competence the industry now faces two problems: how to retain existing skills and competences for the 50 plus years that a plant is operating and how to develop and retain new skills and competences in the areas of decommissioning and radioactive waste management. Different countries are at different stages of the nuclear technology life cycle, a competence that may have declined or be lost in one country may be strong in another. And therein lies one solution to the problems the sector faces - international collaboration. The initiatives such as the NEA Halden project and the Generation IV International Forum lay a ground for quiet optimism that collaboration, information exchange and exchange of personnel continue to be an integral part of the development of nuclear power. Also there is evidence that myriad initiatives are underway in the area of nuclear education and training. Though, national surveys show that still more engineers and scientists having nuclear knowledge are required than are graduating. (author)

  12. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  13. Building Public Trust in Nuclear Power

    International Nuclear Information System (INIS)

    2013-01-01

    Stakeholder involvement is recognized as a crucial process for the success of any nuclear power programme. Failing to effectively engage with stakeholders such as policy- and decision-makers, media, community members, and the public in general can have negative consequences, says Brenda Pagannone, specialist in stakeholder involvement in the IAEA's Nuclear Power Engineering Section. Shaken public trust may lead to delays, and delays are costly for the operator, and the country, and challenging for populations in need of energy. IAEA Member States are increasingly requesting IAEA assistance in their efforts to engage stakeholders. The IAEA is responding to these requests by organizing training and developing guidelines to share expertise and experience and by reviewing national communication strategies. In all of these activities, the IAEA strongly encourages Member States to involve stakeholders throughout the lifecycle of the nuclear power programme. Although each country has specific sets of stakeholders with unique needs and concerns, some principles apply widely.

  14. Nuclear New Build: Insights into Financing and Project Management

    International Nuclear Information System (INIS)

    Horst Keppler, Jan; Cometto, Marco; Kim, Sang-Baik; Sozoniuk, Vladislav; Rothwell, Geoffrey; Thompson, Orme; Savage, Chris; Mancini, Mauro; Leigne, Philippe; Bickford, Erica; Crozat, Matt

    2015-01-01

    Nuclear new build has been progressing steadily since the year 2000, with the construction of 94 new reactors initiated and 56 completed reactors connected to the grid. Among these new reactors are some of the first generation III/III+ reactors of their kind. Drawing on a combination of conceptual analysis, expert opinion and seven in-depth case studies, this report provides policy makers and stakeholders with an overview of the principal challenges facing nuclear new build today, as well as ways to address and overcome them. It focuses on the most important challenges of building a new nuclear power plant, namely assembling the conditions necessary to successfully finance and manage highly complex construction processes and their supply chains. Different projects have chosen different paths, but they nonetheless share a number of features. Financing capital-intensive nuclear new build projects requires, for example, the long-term stabilisation of electricity prices whether through tariffs, power purchase agreements or contracts for difference. In construction, the global convergence of engineering codes and quality standards would also promote both competition and public confidence. In addition, change management, early supply chain planning and 'soft issues' such as leadership, team building and trust have emerged over and again as key factors in the new build construction process. This report looks at ongoing trends in these areas and possible ways forward. (authors)

  15. Passive filtration of air egressing from nuclear containment

    Energy Technology Data Exchange (ETDEWEB)

    Malloy, III, John D

    2017-09-26

    A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

  16. Building a universal nuclear energy density functional

    International Nuclear Information System (INIS)

    Bertsch, G F

    2007-01-01

    This talk describes a new project in SciDAC II in the area of low-energy nuclear physics. The motivation and goals of the SciDAC are presented as well as an outline of the theoretical and computational methodology that will be employed. An important motivation is to have more accurate and reliable predictions of nuclear properties including their binding energies and low-energy reaction rates. The theoretical basis is provided by density functional theory, which the only available theory that can be systematically applied to all nuclei. However, other methodologies based on wave function methods are needed to refine the functionals and to make applications to dynamic processes

  17. 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

  18. Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1977-01-01

    The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

  19. Control of hydrogen concentration in reactor containment buildings by using passive catalytic recombiners

    International Nuclear Information System (INIS)

    Wolff, U.

    1993-01-01

    Severe accidents in nuclear power plants have the potential to generate hydrogen within the reactor containment building in concentrations likely to deflagrate or even detonate. This could endanger the containment integrity. Autocatalytic devices have been developed by the NIS company in Hanau, Germany, to control the hydrogen concentration within the containment. These devices have been tested by the Battelle Institute in Frankfurt, Germany, under conditions relevant to severe accidents. The catalytic device functions as required in a wide band of gas mixtures ranging from inerted conditions with low-hydrogen and/or low-oxygen concentrations up to detonable mixtures. The device starts up quickly, and has a high resistance against catalyst poisons including the effects of oil or cable fires. The device makes a strong contribution to gas mixing in the containment atmosphere. The paper summarizes the development work done and describes the final design of the device. Theoretical tools for analysis and prediction of catalyst performance in containment environments have been developed by the Battelle Institute and the Technical University of Munich. These tools have been verified and validated against experimental data. A phenomenological discussion of accident scenarios is used to explain the functional requirements for the autocatalytic devices in the control of hydrogen. Both the potential for and limitations of such devices for hydrogen control are discussed for large dry containments (PWRs) and for those which are originally inerted (BWRs)

  20. Protective effect of building against nuclear fallout

    International Nuclear Information System (INIS)

    Mueck, K.; Lovranich, E.; Steger, F.

    1995-02-01

    In order to adopt appropriate countermeasures to protect the public in case of a wide-spread contamination after a severe reactor accident, a profound knowledge of the dose to be expected and of the dose reduction to be expected if a specific countermeasure is adopted is required. Since external radiation contributes a major fraction of up to 80 % to the total dose, the interest is focused on the reduction of external dose by staying indoors. For this purpose measurements of the dose reduction by external radiation were performed in 40 residential rooms in Vienna which were chosen according to their fraction of the Viennese building structure. They were performed by measurement of the reduction of the gamma-flux in the interior of buildings compared to the gamma-flux outside. Measurements were carried out by a HPGe-detector which was positioned 1 m above ground. The observed dose reduction factors amounted to 0.013 on the average (protection factor 78.7 ± 49.7), with the protection factors of the investigated building types ranging from 6.5 (single house in garden) to 122.9 (edicifices of the turn of the century). The observed protection factors do not include the dose reduction by dense arrangement of buildings in urban areas which has to be considered separately. The dose reduction due to different radionuclide mixtures after severe reactor accidents was also investigated. Factors similar to that observed for Cs-137 were found. The maximum deviations amounted only 12 % and therefore may be neglected in the consideration of protective measures. Additional measurements were performed with the same detector shielded by lead on all sides to determine the contribution of the interior contamination. A contribution of between < 10 % to 50 % to the external dose in the interior of buildings was observed. Measures to reduce the inner contamination in rooms (closing of windows and doors, cleaning upon entering the building interior) in case of an event are therefore

  1. Building a Universal Nuclear Energy Density Functional

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Joe A. [Michigan State Univ., East Lansing, MI (United States); Furnstahl, Dick; Horoi, Mihai; Lust, Rusty; Nazaewicc, Witek; Ng, Esmond; Thompson, Ian; Vary, James

    2012-12-30

    During the period of Dec. 1 2006 – Jun. 30, 2012, the UNEDF collaboration carried out a comprehensive study of all nuclei, based on the most accurate knowledge of the strong nuclear interaction, the most reliable theoretical approaches, the most advanced algorithms, and extensive computational resources, with a view towards scaling to the petaflop platforms and beyond. The long-term vision initiated with UNEDF is to arrive at a comprehensive, quantitative, and unified description of nuclei and their reactions, grounded in the fundamental interactions between the constituent nucleons. We seek to replace current phenomenological models of nuclear structure and reactions with a well-founded microscopic theory that delivers maximum predictive power with well-quantified uncertainties. Specifically, the mission of this project has been three-fold: First, to find an optimal energy density functional (EDF) using all our knowledge of the nucleonic Hamiltonian and basic nuclear properties; Second, to apply the EDF theory and its extensions to validate the functional using all the available relevant nuclear structure and reaction data; Third, to apply the validated theory to properties of interest that cannot be measured, in particular the properties needed for reaction theory.

  2. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

    1994-01-01

    Research is being conducted by ORNL under US Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques. assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants

  3. PRTR/309 building nuclear facility preliminary

    International Nuclear Information System (INIS)

    Cornwell, B.C.

    1994-01-01

    The hazard classification of the Plutonium Recycle Test Reactor (PRTR)/309 building as a ''Radiological Facility'' and the office portions as ''Other Industrial Facility'' are documented by this report. This report provides: a synopsis of the history and facility it's uses; describes major area of the facility; and assesses the radiological conditions for the facility segments. The assessment is conducted using the hazard category threshold values, segmentation methodology, and graded approach guidance of DOE-STD-1027-92

  4. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-08-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. The program encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  5. Crash testing of nuclear fuel shipping containers

    International Nuclear Information System (INIS)

    Jefferson, R.M.; Yoshimura, H.R.

    1977-12-01

    In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. This program, which began in 1975, encompasses the following distinct major efforts. The first of these utilizes computational methods for predicting the effects of the accident environment and, subsequently, to calculate the damage incurred by a container as the result of such an accident. The second phase involves the testing of 1 / 8 -scale models of transportation systems. Through the use of instrumentation and high-speed motion photography, the accident environments and physical damage mechanisms are studied in detail. After correlating the results of these first two phases, a full scale event involving representative hardware is conducted. To date two of the three selected test scenarios have been completed. Results of the program to this point indicate that both computational techniques and scale modeling are viable engineering approaches to studying accident environments and physical damage to shipping casks

  6. Nuclear component design ontology building based on ASME codes

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2005-01-01

    The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)

  7. Nuclear Energy Development and New Build Expansion

    International Nuclear Information System (INIS)

    Stosic, Z. V.

    2012-01-01

    Early afternoon on March 11th, 2011, a devastating earthquake hit Japan, causing a powerful tsunami which had catastrophic consequences in the Tohoku District. A nuclear accident followed with core meltdowns at the Fukushima Daiichi NPPs (Nuclear Power Plants) at an unprecedented scale and over a lengthy period of time. The findings so far suggest that the insufficient design for tsunamis of the reactor units was responsible for the accident that occurred in the Japanese Fukushima Daiichi NPP. Thus the accident does not fall into the category of residual risk; rather it was due to the fact that the basic design for external impact was insufficient in this case. This is why the design and the safety concept of NPPs around the world had to be reviewed with respect to possible improvement potential. The impact of the Tohoku natural disaster is present not only in Japan but world-wide. The context post-Fukushima creates new challenges, but nuclear perspectives remain solid despite shaken public acceptance and the fundamentals driving nuclear role in sustainable energy mix remain. These are: GROWING DEMAND: Need for new capacity is unchanged to meet growing energy demand (multiplied by two in overall consumption and an 80% increase in global electricity consumption by 2050); REDUCTION OF CO 2 EMISSIONS: Although 50% of world electricity today is generated from burning coal, combating climate change remains a priority and greenhouse gas emissions are to be cut by half by 2050; SECURITY OF SUPPLY: Need for an increased security of supply in a changing geopolitical environment; FOSSIL ENERGY: Fossil resources are dwindling, remain uncertain and are volatile in prices; COMPETITIVENESS: Nuclear remains one of the most competitive low-carbon energy sources and will remain an important option for many countries for a sustainable energy mix. To supply seven billion people (nine billion in 2030) with secure energy needs infrastructure development. This means huge investments

  8. Phonons as building blocks in nuclear structure

    International Nuclear Information System (INIS)

    Silvestre-Brac, B.

    1980-01-01

    The structure of a nuclear system in terms of eigenmodes (phonons) of subsystems is investigated in three different approaches. In the frame of nuclear field theory the three identical particle system is analysed and the elimination of spurious states due to the violation of the Pauli principle is emphasized. In terms of weak coupling, a new approach of the shell model is proposed which is shown to be rapidly convergent with the number of basis vectors. Applications of three particle systems in the lead region are made. Lastly, a microscopic multiphonon theorie of collective K=0 states in deformed nuclei based on a Tamm Dancoff phonon is developed. The role of the Pauli principle as well as comparisons with boson expansion methods are deeply analysed [fr

  9. Decommissioning of building part of nuclear power plant

    International Nuclear Information System (INIS)

    Sochor, R.

    1988-01-01

    The characteristics are discussed using literature data of building work during decommissioning or reconstruction of nuclear power plants. The scope of jobs associated with power plant decommissioning is mainly given by the size of contaminated parts, intensity of radioactivity, the volume of radioactive wastes and the possible building processes. Attention is devoted to the cost of such jobs and the effect of the plant design on cost reduction. (Z.M.). 6 refs

  10. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  11. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    Swenson, C.E.

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ''Technical Baseline and Updated Cost Estimate for the Canister Storage Building'', dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995

  12. Guidelines for Assessment and Abatement of Asbestos-Containing Materials in Buildings.

    Science.gov (United States)

    Pielert, James H.; Mathey, Robert G.

    This report presents guidelines, based on available information, for the assessment and abatement of asbestos-containing materials in buildings. Section 1 provides background information on the history and use of asbestos-containing products in buildings, the characteristics of asbestos fibers, products and materials containing asbestos, and…

  13. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  14. British Nuclear Fuels plc's effluent plant services building

    International Nuclear Information System (INIS)

    Williams, L.

    1990-01-01

    The new Effluent Plant Services building (EPSB) on the Sellafield Nine Acre Site was built by Costain Engineering Limited for British Nuclear Fuels Limited. The EPSB is dedicated to a new generation of nuclear waste treatment plants, aimed at reducing discharges into the Irish Sea and other environmental impacts by removing actinides from liquid effluents and decontaminating waste solvents. This article describes the design, construction and operation of the plant. (UK)

  15. Development of human resource capacity building assistance for nuclear security

    International Nuclear Information System (INIS)

    Nakamura, Yo; Noro, Naoko

    2014-01-01

    The Integrated Support Center for Nuclear Nonproliferation and Nuclear Security (ISCN) of the Japan Atomic Energy Agency (JAEA) has been providing nuclear security human resource development projects targeting at nuclear emerging countries in Asia in cooperation with the authorities concerned including the Sandia National Laboratory (SNL) and the International Atomic Energy Agency (IAEA). In the aftermath of the attacks of Sept. 11, the threat of terrorism was internationally recognized and thus the human resource capacity building is underway as an urgent task. In order to responding to emerging threats, the human resource capacity building that ISCN has implemented thus far needs to be multilaterally analyzed in order to develop more effective training programs. This paper studies ISCN's future direction by analyzing its achievements, as well as introduces the collaborative relationships with SNL that contributes to the reflection and maintenance of international trends for the contents of nuclear security training, the nuclear security enhancement support with which Japan is to provide nuclear emerging countries in Asia, and the achievements of the nuclear security training program that ISCN implemented. (author)

  16. Preparation of the Regulatory Infrastructure for the New Nuclear Build

    International Nuclear Information System (INIS)

    Cimesa, Sinisa; Persic, Andreja; Vrankar, Leopold; Stritar, Andrej

    2011-01-01

    Slovenia is seriously considering building a new nuclear power plant. The Slovenian Nuclear Safety Administration (SNSA) is very much aware of the complexity of such a project as well as of the fact that at the moment the SNSA does not have sufficient resources for licensing and overseeing the design, construction and operation of the possible new plant. Likewise, the question arises whether technical support organizations which support SNSA in supervising the existing Krsko nuclear power plant have sufficient capacity. Therefore SNSA established a special project team with the task to prepare the Administration for the possible start of the new nuclear build. In the beginning of 2009, the project team prepared the analysis of licensing process, which is basically an overview of spatial planning, construction and nuclear safety regulation processes. The purpose of the review of the whole process, from spatial planning to the issuance of the operating license, was to identify phases which will require most effort. The next step was to set the strategy for the review process as well as to analyze and establish the basis for resource demands needed for SNSA's and other stakeholders involvements and decision making in the process. This will enable SNSA to establish a qualified and effective infrastructure for a possible new nuclear build. (authors)

  17. Periodic Safety Review of Tendon Pre-stress of Concrete Containment Building for a CA U-Type clear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Kwang Ho; Lim, Woo Sang [Korea Hydro and clear Power Co., Daejeon (Korea, Republic of)

    2009-10-15

    Generally, as the tendon pre-stress of concrete containment buildings at nuclear power plants decreases as time passes due to the concrete creep, concrete shrinkage and the relaxation of tendon strands, the tendon pre-stress must secure the structural integrity of these buildings by maintaining its value higher than that of the designed pre-stress during the overall service life of the nuclear power plants. Moreover, if necessary, the degree of tendon pre-stress must also guarantee the structural integrity of concrete containment buildings over their lifetimes. This paper evaluated the changes in the tendon pre-stress of a concrete containment building subject to time-limited aging as an item in a periodic safety review (PSR) of Wolsong unit 1, a CANDU-type nuclear power plant to ensure that the structural integrity can be maintained until the next PSR period after the designed lifetime.

  18. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  19. Pressure release in containments of nuclear power stations

    International Nuclear Information System (INIS)

    Pauli, W.; Pellaud, B.; Saitoh, A.

    1992-01-01

    In France, Germany, Sweden and Switzerland, the licensing authorities have decided to equip nuclear reactor containments with a filter venting system to ensure survival of the containment after postulated severe nuclear accidents. This is a curious paradox. For years, the established wisdom was unambiguously 'Keep the containment tight. It's the ultimate barrier.' Three Mile Island seemed to prove the point. Yet, an old mechanical engineer's rule is 'Every pressure vessel must have a safety valve.' Filtered containment venting attempts to reconcile these two conflicting objectives by allowing a filtered pressure relief after an accident, in order to prevent containment failure due to overpressure, while keeping the release within acceptable limits. Achieving this dual objective is a matter of proper timing, i.e. pressure relief, not too early, not too late. (author)

  20. Device for separating, purifying and recovering nuclear fuel material, impurities and materials from impurity-containing nuclear fuel materials or nuclear fuel containing material

    International Nuclear Information System (INIS)

    Sato, Ryuichi; Kamei, Yoshinobu; Watanabe, Tsuneo; Tanaka, Shigeru.

    1988-01-01

    Purpose: To separate, purify and recover nuclear fuel materials, impurities and materials with no formation of liquid wastes. Constitution: Oxidizing atmosphere gases are introduced from both ends of a heating furnace. Vessels containing impurity-containing nuclear fuel substances or nuclear fuel substance-containing material are continuously disposed movably from one end to the other of the heating furnace. Then, impurity oxides or material oxides selectively evaporated from the impurity-containing nuclear fuel substances or nuclear fuel substance-containing materials are entrained in the oxidizing atmosphere gas and the gases are led out externally from a discharge port opened at the intermediate portion of the heating furnace, filters are disposed to the exit to solidify and capture the nuclear fuel substances and traps are disposed behind the filters to solidify and capture the oxides by spontaneous air cooling or water cooling. (Sekiya, K.)

  1. Nuclear power costs in the build, operate, transfer approach

    International Nuclear Information System (INIS)

    Aybers, M.N.; Sahin, B.

    1990-01-01

    The costs of nuclear power are discussed with special reference to the economic problems faced by developing countries, and the relative merit of a new accounting approach, viz., the build, operate, transfer contract model, which was proposed in Turkey for the Akkuyu nuclear power project, is illustrated. In this context, the general methodology of calculating nuclear power costs is summarized and a capital cost analysis for a 986 MW pressurized water reactor plant is given in terms of constant monetary units for the above contract model and the turnkey contract model. Adjustment of the costs taking into account regional conditions such as inflation and higher interest rates is also indicated. (orig.) [de

  2. Semipalatinsk nuclear test site: History of building and function

    International Nuclear Information System (INIS)

    Sergazina, G.M.; Balmukhanov, S.B.

    1999-01-01

    A vast materials on history of Semipalatinsk nuclear test site creation and it building and function are presented. Authors with big reliability report one page of Kazakhstan's history. In steppe on naked place thousands of soldiers and officers, construct and military specialists have built the nuclear site on which during 40 years were conducting nuclear tests . Prolonged chronic radiation on population living near by site results to tragedy which is confessed by General Assembly of United Nations. In the book aspects of test site conversion and rehabilitation of injured population are considered. The book consists of introduction, three chapters and conclusion. The book is intended to wide circle of readers. (author)

  3. Building, measuring and improving public confidence in the nuclear regulator

    International Nuclear Information System (INIS)

    2006-01-01

    An important factor for public confidence in the nuclear regulator is the general public trust of the government and its representatives, which is clearly not the same in all countries. Likewise, cultural differences between countries can be considerable, and similar means of communication between government authorities and the public may not be universally effective. Nevertheless, this workshop identified a number of common principles for the communication of nuclear regulatory decisions that can be recommended to all regulators. They have been cited in particular for their ability to help build, measure and/or improve overall public confidence in the nuclear regulator. (author)

  4. Stresses on nuclear power plant buildings by extraordinary events

    International Nuclear Information System (INIS)

    Woelfel, E.

    1977-01-01

    Nuclear power plant buildings must be functional to such an extend that even after the occurence of extraordinary events (earthquake, airoplane crash, gas cloud explosion), the reactor can be safety shut off, in order to avoid danger from the nuclear power plant. Evidence for this can only be given by calculations which shall meet the following requirements: The calculation results shall be safe and reliable. The calculation effort shall match the realizable accuracy. The calculation shall lead to an economical determination. An example of ascertainment of nuclear power plants in regard to earthquakes, shows the difficulties standing against a fulfillment of these requirements. (orig.) [de

  5. Float level switch for a nuclear power plant containment vessel

    International Nuclear Information System (INIS)

    Powell, J.G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures

  6. Float level switch for a nuclear power plant containment vessel

    Science.gov (United States)

    Powell, James G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  7. Capacity building in the IAEA Action Plan on Nuclear Safety

    International Nuclear Information System (INIS)

    Caruso, Gustavo

    2014-01-01

    Conclusion and Future Challenges: • Momentum on Nuclear Safety; • To continue strengthening, developing, maintaining and implementing capacity building programmes, including education, training and exercises at the national, regional and international levels; • To ensure sufficient and competent human resources necessary to assume responsibility for safety; • To incorporate lessons learned from the accident based on the IEMs and IAEA Fukushima Report

  8. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    Reuter, H.R.; Chang-Lo, P.L.C.; Pfeifer, B.W.; Shah, G.H.; Whitcraft, J.S.

    1975-02-01

    A discussion of the techniques and procedures used for the design of prestressed concrete nuclear reactor containment structures is presented. A physical description of Bechtel designed containment structures is presented. The design bases and load combinations are given for anticipated conditions of service. Reference design documents which include industry codes, specifications, AEC Regulatory Guides, Bechtel Topical Reports and additional criteria as appropriate to containment design are listed. Stepwise procedures typically followed by Bechtel for design of containments is discussed and design examples are presented. A description of currently used analytical methods and the practical application of these methods for containment design is also presented. The principal containment construction materials are identified and codes of practice pertaining to construction procedures are listed. Preoperational structural testing procedures and post-operational surveillance programs are furnished along with results of tests on completed containment structures. (U.S.)

  9. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  10. Damage assessment of nuclear containment against aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, Mohd Ashraf, E-mail: iqbal_ashraf@rediffmail.com; Sadique, Md. Rehan, E-mail: rehan.sadique@gmail.com; Bhargava, Pradeep, E-mail: bhpdpfce@iitr.ac.in; Bhandari, N.M., E-mail: nmbcefce@iitr.ac.in

    2014-10-15

    Highlights: • Damage assessment of nuclear containment is studied against aircraft crash. • Four impact locations have been identified at the outer containment shell. • The mid of the total height has been found to be most vulnerable location. • The crown of dome has been found to be the strongest location. • Phantom F4 caused more localized and severe damage compared to other aircrafts. - Abstract: The behavior of nuclear containment structure has been studied against aircraft crash with an emphasis on the influence of strike location. The impact locations identified on the BWR Mark III type nuclear containment structure are mid-height, junction of dome and cylinder, crown of dome and arc of dome. The containment at each of the above locations has been impacted normally by Phantom F-4, Boeing 707-320 and Airbus A320 aircrafts. The loading of the aircraft has been assigned through the corresponding reaction-time response curve. ABAQUS/Explicit finite element code has been used to carry out the three-dimensional numerical simulations. The concrete damaged plasticity model was used to simulate the behavior of concrete while the behavior of steel reinforcement was incorporated using the Johnson–Cook elasto-viscoplastic material model. The mid-height of containment has been found to experience most severe deformation against each aircraft. Phantom F4 has been found to be most disastrous at each location. The results have been compared with those of the available studies with respect to the containment deformation.

  11. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  12. Aging characteristics of containment building and sensitivity on ultimate pressure capacity

    International Nuclear Information System (INIS)

    Seo, Jeong Moon; Choun, Young Sun; Choi, In Kil; Ha, Jae Joo

    1998-03-01

    For the reliable safety assessment of the containment building, structural and material conditions can be investigated in detail and pertinent assessment technologies have to be established. Also, an understanding on the aging-related degradations for the construction materials is required to predict long-term structural safety of the containment building. For the development of reliable aging prediction models, an extensive data base system related to aging properties of the containment building has to be prepared. The objectives of this research are to develop aging models representing long-term degradation of materials and a structural performance assessment program for containment building considering aging-related degradation. According to the results of sensitivity analysis, as the mechanical properties of the constituent materials degrade, the ultimate pressure capacity of containment building may decrease and severe damage may occur around the mid-level of the containment wall. (author). 28 refs., 11 tabs., 36 figs

  13. Innovative nuclear power plant building arragement in consideration of decommissioning

    International Nuclear Information System (INIS)

    Choi, Won Jun; Roh, Myung Sub; Kim, Chang Lak

    2017-01-01

    A new concept termed the Innovative Nuclear Power Plant Building Arrangement (INBA) strategy is a new nuclear power plant building arrangement method which encompasses upfront consideration of more efficient decommissioning. Although existing decommissioning strategies such as immediate dismantling and differed dismantling has the advantage of either early site restoration or radioactive decommissioning waste reduction, the INBA strategy has the advantages of both strategies. In this research paper, the concept and the implementation method of the INBA strategy will be described. Two primary benefits will be further described: (1) early site restoration; and (2) radioactive waste reduction. Several other potential benefits will also be identified. For the estimation of economic benefit, the INBA strategy, with two primary benefits, will be compared with the immediate dismantling strategy. The effect of a short life cycle nuclear power plant in combination with the INBA strategy will be reviewed. Finally, some of the major impediments to the realization of this strategy will be discussed

  14. Innovative nuclear power plant building arragement in consideration of decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Won Jun; Roh, Myung Sub; Kim, Chang Lak [Dept. of Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2017-04-15

    A new concept termed the Innovative Nuclear Power Plant Building Arrangement (INBA) strategy is a new nuclear power plant building arrangement method which encompasses upfront consideration of more efficient decommissioning. Although existing decommissioning strategies such as immediate dismantling and differed dismantling has the advantage of either early site restoration or radioactive decommissioning waste reduction, the INBA strategy has the advantages of both strategies. In this research paper, the concept and the implementation method of the INBA strategy will be described. Two primary benefits will be further described: (1) early site restoration; and (2) radioactive waste reduction. Several other potential benefits will also be identified. For the estimation of economic benefit, the INBA strategy, with two primary benefits, will be compared with the immediate dismantling strategy. The effect of a short life cycle nuclear power plant in combination with the INBA strategy will be reviewed. Finally, some of the major impediments to the realization of this strategy will be discussed.

  15. Innovative Nuclear Power Plant Building Arrangement in Consideration of Decommissioning

    Directory of Open Access Journals (Sweden)

    Won-Jun Choi

    2017-04-01

    Full Text Available A new concept termed the Innovative Nuclear Power Plant Building Arrangement (INBA strategy is a new nuclear power plant building arrangement method which encompasses upfront consideration of more efficient decommissioning. Although existing decommissioning strategies such as immediate dismantling and differed dismantling has the advantage of either early site restoration or radioactive decommissioning waste reduction, the INBA strategy has the advantages of both strategies. In this research paper, the concept and the implementation method of the INBA strategy will be described. Two primary benefits will be further described: (1 early site restoration; and (2 radioactive waste reduction. Several other potential benefits will also be identified. For the estimation of economic benefit, the INBA strategy, with two primary benefits, will be compared with the immediate dismantling strategy. The effect of a short life cycle nuclear power plant in combination with the INBA strategy will be reviewed. Finally, some of the major impediments to the realization of this strategy will be discussed.

  16. Nuclear power stations. Information paper no. 1. Controls on the building and running of nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    1981-09-01

    Controls and constraints which govern the development and running of nuclear power stations are briefly examined. Government policy, permission to build, authority to start building, site acquisition building, running and public opinion are briefly discussed.

  17. Local competence building and public information in European nuclear territories

    International Nuclear Information System (INIS)

    Martell, Meritxell; Vila D'Abadal, Maria

    2010-01-01

    Opinion polls and case study analysis show that there is a strong public demand for more participation in decision-making processes relating to the environment and nuclear issues specifically (IGNA, 2007; EC DG TREN 2008). In this regard, the implementation of transparency and participation at local level in municipalities hosting nuclear facilities is being claimed by local representatives over Europe. The pilot project 'Local Competence Building and Public Information in European Nuclear Territories' (2007-2008) was promoted by the Group of European Municipalities with Nuclear Facilities (GMF) and partly financed by the Directorate General for Energy and Transport (DG TREN). The project aims at supporting the European Commission (EC) in the process of harmonising national practices in the field of governance in nuclear territories. The objectives of this pilot project were two-fold: firstly, to identify local good practices at European level with regards to governance in the nuclear field and, secondly, to provide a methodological framework to recommend good practices and establish a set of indicators. The results of the project are briefly explained under the five dimensions of local governance on nuclear matters: access to environmental information, participation, access to justice, competence building and local development. Firstly, access to information on nuclear issues is broadly acknowledged by local representatives to be of relevant importance in increasing public knowledge on nuclear issues, especially on nuclear safety and radiation protection. Secondly, most local representatives feel that the current participatory level on nuclear issues in municipalities is too low and they would appreciate a greater involvement of local communities in decision-making processes on nuclear issues through local commissions or other mechanisms. Thirdly, according to local authorities, access to justice should be provided not only for the public but also for themselves

  18. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

    The containment is a structural envelope which completely surrounds the nuclear reactor system and is designed to confine the radioactive releases in case of an accident. This report summarises the work of an NEA Senior Group of Experts who have studied the potential role of containment in accidents exceeding design specifications (so-called severe accidents). Some possibilities for enhancing the ability of plants to reduce the risk of significant off-site consequences by appropriate management of the acident have been examined

  19. Permanent cavity seal ring for a nuclear reactor containment arrangement

    International Nuclear Information System (INIS)

    Swidwa, K.J.; Salton, R.B.; Marshall, J.R.

    1990-01-01

    This patent describes a nuclear reactor containment arrangement. It comprises: a reactor pressure vessel which thermally expands and contracts during cyclic operation of the reactor, the vessel having a peripheral wall and a horizontally outwardly extending flange thereon; a containment wall having a shelf, the wall spaced from and surrounding the peripheral wall of the reactor pressure vessel defining an annular expansion gap therebetween, and an annular ring seal extending across the annular expansion gap to provide a water-tight seal therebetween

  20. Nuclear reactor melt-retention structure to mitigate direct containment heating

    Science.gov (United States)

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  1. Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

    International Nuclear Information System (INIS)

    Bobovnik, G.; Kljenak, I.

    2001-01-01

    Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

  2. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  3. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  4. Study of vibration analysis for nuclear reactor building

    International Nuclear Information System (INIS)

    Hirashima, Shin-ichi

    1978-01-01

    The mutual interference between the contiguous buildings with separate foundations and also that between the outer wall under the ground and the foundation bottom of the building were taken into consideration for the vibration analysis with spring-mass system. For two contiguous foundations of buildings it was attempted to represent the static mutual interference by a spring-mass system model. The theoretical analysis formulas are shown for the combination of the vertical movement and rocking motion, and for the interfering forces between the foundation and the outer wall of a building. The method of extending the model to dynamic one is explained. Several spring constants utilized in the analysis were obtained, for example, for mutual interference springs regarding vertical motion, mutual interfering springs for the foundation and the outer wall of a building and the mutual interference springs concerning horizontal movement. These models and analysis were applied to the BWR-MARK II-1100 MW nuclear reactor building and the contiguous turbine building. The structures and level relations of two buildings are shown, and the spring-mass system model for these buildings is expressed. The masses of about 20, the weights, the rotating inertia, the sectional moment of inertia, the spring constant and the damping coefficient for each mass are tabulated. As the results, the peak displacements occur at 2.556 Hz, 6.918 Hz, 10.43 Hz and 13.85 Hz. The damping coefficient is large and about 10 - 30% at the lower order modes. The calculated and the measured vibration characteristics for the BWR plant buildings are not much different, and this spring-mass system model is verified to be adequate. (Nakai, Y.)

  5. Building public confidence in the world's nuclear industry

    International Nuclear Information System (INIS)

    Duncan, C.D.

    1996-01-01

    Public confidence in the nuclear industry requires two things, which are trust and understanding. Trust is an emotional response based upon an instinctive reaction. Understanding, on the other hand, is an intellectual response based upon facts. To gain public confidence, both of these levels must be communicated and proactive strategies must be implemented to do this. To achieve this objective will require confidence and courage in communication programs. Each company operating in the nuclear sector must be proactive in building its individual reputation and must not retreat from controversy. Similarly, each industry body must continue the Herculean task of building understanding. The nuclear industry has powerful arguments. ICI, BP or Ford did not achieve their licences to operate by keeping their heads down, they achieved their current market positions by building a positive corporate reputation within their respective industrial contexts over many decades. In order to achieve a similar position for the nuclear industry and the companies, their examples must be followed. If it is continued to 'keep the heads down' in the trenches, public opinion will surely bury within it. (G.K.)

  6. Aircraft crash upon outer containment of nuclear power plant

    International Nuclear Information System (INIS)

    Abbas, H.; Paul, D.K.; Godbole, P.N.; Nayak, G.C.

    1996-01-01

    In this paper, analysis of an aircraft crash upon an outer containment of a nuclear power plant is presented. The effect of target yielding is considered simultaneously by calculating the reaction time in a time marching scheme. The concrete model employed is capable of predicting the cracking and yielding. The response for different cracking strains and different locations of aircraft strike for different aircraft has been studied. Critical location of aircraft strike for the containment has been investigated. The analytical procedure and the material model used are found to be capable of representing the aircraft impact response of the containment structure. (orig.)

  7. Impact of risk analysis on ITER nuclear buildings layout

    International Nuclear Information System (INIS)

    Sabathe, Laurent; Lignini, Franck; Rodriguez-Rodrigo, Lina; Uzan-Elbez, Joelle; Girard, Jean-Philippe

    2007-01-01

    The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of the risks induced by normal and incidental or accidental conditions: - radiological exposure to tritium, and activated products → radiological zoning; - exposure to beryllium (vacuum vessel blankets) → beryllium zoning; - exposure to electromagnetic fields during plasma shots → electromagnetic zoning; - fire hazard: fire compartmentalization; ·explosion hazard: anti-blast zoning; - waste treatment → waste zoning. Prevention of common mode failure due to internal hazards (fire, flooding, load drop...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer has also taken into account access requirements to equipments and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows meeting the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications are been studied in the tritium building: utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from

  8. Impact of risk analysis on ITER nuclear buildings internal layout

    International Nuclear Information System (INIS)

    Lignini, F.; Sabathe, L.; Rodriguez-Rodrigo, L.; Uzan-Elbez, J.; Girard, J.-P.

    2006-01-01

    The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of risk induced by normal and incidental or accidental conditions : - radiological exposure to tritium, and activated products → radiological zoning - exposure to beryllium (vacuum vessel blankets) → beryllium zoning - exposure to electromagnetic fields during plasma shots → electromagnetic zoning - fire hazard : fire compartmentalisation - explosion hazard → anti-blast zoning - waste treatment → waste zoning Prevention of common mode failure due to internal hazards (fire, flooding, load drop ...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer also has to take into account access requirements to equipment and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows to meet the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications were included in the tritium building : utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from the

  9. Consideration on hydrogen explosion scenario in APR 1400 containment building during small breakup loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweonha, E-mail: khpark@kmou.ac.kr [Division of Mechanical & Energy Systems Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of); Khor, Chong Lee, E-mail: itachi_829@hotmail.com [Department of Mechanical Engineering, Korea Maritime University, Dongsam-dong, Yeongdo-gu, Busan 606-791 (Korea, Republic of)

    2015-11-15

    Highlights: • Hydrogen behavior in the containment building of APR1400 nuclear plant up to 15 h after the failure happened. • The risk of hydrogen explosion largely depends on the combination of air, hydrogen and steam in the containment. • Hydrogen explosion risk at different locations in the containment was analyzed. - Abstract: This paper describes the analytical result of the potential risk of hydrogen gas up to 15 h after the failure takes place. The major cause of the disaster occurred in Fukushima Daiichi nuclear reactor was the detonation of accumulated hydrogen in the containment by highly increased reactor core temperatures after the failure of the emergency cooling system. The hydrogen risk should be considered in severe accident strategies in current and future NPPs. A hydrogen explosion scenario is proposed. Hydrogen is accumulated on top of the dome during the hydrogen release period. At this point, there are no risk of explosion due to the steam that resides in upper part of the dome. As the hydrogen concentration increase, substantial amount of steams are released. Subsequently, hydrogen is forced into the lower part of the building with high air density—small explosion and dormant steam condensation phase are possible. The light hydrogen rises up slowly with air, gathering on top of the building with high air density. Massive hydrogen explosion is anticipated upon ignition at this stage.

  10. Concrete containment vessels (CCV) for nuclear power plants, (1)

    International Nuclear Information System (INIS)

    Ibe, Yukimi; Kitajima, Masatake

    1977-01-01

    Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

  11. Nuclear safety review for qualification of class 1E motor inside containment for nuclear power stations

    International Nuclear Information System (INIS)

    Li Shixin; Wu Qi; Zhang Yunbo; Wu Caixia

    2013-01-01

    In nuclear power plants with pressurized water reactors, the review for class 1E motor inside containment qualification process and documents is an important aspect of nuclear safety equipment review, and the reviewers should make evaluations for the qualification test results in terms of the compliance with standard and regulation, and the consistency with inside containment environment. Firstly, this paper introduces the qualification test of class 1E motor inside containment for nuclear power generating stations, such as aging test and design-basis-event test. Second, there is a discussion about typical problems of review. At last, comparison of IEEE334 with RCC-E is conducted and explored. (authors)

  12. A regulatory view of nuclear containment on UK licensed sites

    International Nuclear Information System (INIS)

    Bradford, P.M.; McNair, I.J.

    1997-01-01

    Members of the UK regulatory body, HM Nuclear Installations Inspectorate (NII) have previously presented conference papers and official reports which have dealt separately with either reactor applications or chemical plant applications. The objective of this paper is to draw together a brief overview of the role of containment in protecting against potential radiological and related hazards, and to describe the factors which influence the NII's assessment of containment safety cases. It draws upon the NII's experience of regulating many types of nuclear facility, from those designed in the late 1940s through to the modern plants, such as Sizewell 'B' and THORP. The paper reviews the legislative and regulatory background within which the facilities exist and are operated. Finally, the paper reviews recent, ongoing and planned research in the field of containment, which has been designed to behave under challenge. (author)

  13. Aerosol behavior in the reactor containment building during severe accident

    International Nuclear Information System (INIS)

    Berthion, Y.; Lhiaubet, G.; Gauvain, J.

    1984-07-01

    Thermohydraulic behavior inside a PWR containment during severe accident depends on decay heat transferred to the sump water by aerosol gravitational settling and deposition. Conversely, aerosol behavior depends on thermal hydraulic conditions, especially atmosphere moisture for soluble aerosol GsI, and CsOH. Therefore, a small iterative procedure between thermo-hydraulic and aerosol calculations has been performed in order to evaluate the importance of this coupling between the two phenomena. In this paper, it is shown that with this procedure and using our codes JERICHO, RICOCHET and AEROSOLS/B1, the steam condensation on aerosols is an important phenomenon for a correct estimation of the attenuation factor of the suspended mass of aerosols in the airborne of the containment. Then, we have a more realistic assessment of the source term released by the containment

  14. Containment for low temperature district nuclear-heating reactor

    International Nuclear Information System (INIS)

    He Shuyan; Dong Duo

    1992-03-01

    Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

  15. Collecting and recirculating condensate in a nuclear reactor containment

    International Nuclear Information System (INIS)

    Schultz, T.L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures

  16. Collecting and recirculating condensate in a nuclear reactor containment

    Science.gov (United States)

    Schultz, Terry L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  17. Build of virtual instrument laboratory related to nuclear species specialized

    International Nuclear Information System (INIS)

    Shan Jian; Zhao Guizhi; Zhao Xiuliang; Tang Lingzhi

    2009-01-01

    As rapid development of specialized related to nuclear science,the requirement of laboratory construct is analyzed in this article at first, One total conceive, One scheme deploy soft and hardware,three concrete characteristics targets and five different phases of put in practice of virtual instrument laboratory of specialized related to nuclear science are suggest in the paper,the concrete hardware structure and the headway of build of virtual instrument laboratory are described,and the first step effect is introduced.Lastly,the forward target and the further deliberateness that the virtual instrument laboratory construct are set forth in the thesis. (authors)

  18. Overview of nuclear new build projects and global perspective

    Energy Technology Data Exchange (ETDEWEB)

    Poncelet, Jean-Pol [FORATOM, Brussels (Belgium)

    2017-05-15

    Nuclear power is an important source for electricity production in Europe: today 131 reactors are operated in 14 EU Member States, delivering 28 % of the European power and one half of its low-carbon electricity. The turnover of the sector is about 70 billion Euro and there are about 800,000 high qualified jobs. Worldwide the capacities of nuclear power are extending. New build activities are moving to the Eastern countries. Today, the whole electricity market in Europe is characterised by uncertainties for all investments due to political market interventions. A common European energy policy does not appear to exist.

  19. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.; Graves, H.L. III; Norris, W.E.

    1996-01-01

    Research is being conducted by Oak Ridge National Laboratory under US nuclear regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a structural materials information center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (orig.)

  20. Aging management of containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1994-01-01

    Research is being conducted by Oak Ridge National Laboratory under U.S. Nuclear Regulatory Commission sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the US-NRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of nondestructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. (author). 29 refs., 2 figs

  1. EDF Energy Nuclear New Build: Lessons Learned in Knowledge Management

    International Nuclear Information System (INIS)

    Sachar, M.; Borlodan, G.

    2016-01-01

    Full text: EDF Energy Nuclear New Build (NNB) is building two EPR reactors at Hinkley Point C in Somerset in the United Kingdom that will provide reliable, low carbon electricity to meet approximately 7% of the UK’s electricity needs. The Hinkley Point C project is well advanced. It has achieved planning consent, design approval for the EPR reactor and a nuclear site license. There is a well-developed supply chain with identified preferred bidders who are already heavily involved in construction planning. Training for needed skills is underway and industrial agreements with trade unions are in place. NNB has the unique opportunity to set Knowledge Management behaviours, culture, and standards for the Hinkley Point C project from project inception instead of working to change them, such as on an operational site. (author

  2. Assessment of seismic damages in nuclear power plant buildings

    International Nuclear Information System (INIS)

    Corsanego, A.; DelGrosso, A.; Ferro, G.

    1987-01-01

    Performance of nuclear power plant sites, buildings and components is in today's practice continuously evaluated by means of monitoring systems composed by a variety of instruments, allowing records of the most significant behavioral parameters to be gathered by electronic data acquisition equipment. A great emphasis has been devoted in recent years to the development of ''intelligent'' monitoring systems able to perform interpretation of the response of structures and components automatically, only requiring human intervention and sophisticated data processing techniques when degradation of the safety margins is likely to have been produced. Such computerized procedures can be formulated through logic or algorithmic processes and normally are consistently based upon simplified, heuristic behavioral models and probabilistic reasoning schemes. This paper is devoted to discuss the development of an algorithmic procedure intended for automatic, real-time interpretation of the recorded response of nuclear power plant buildings and foundations during seismic events

  3. Earthquake response of nuclear reactor buildings deeply embedded in soil

    International Nuclear Information System (INIS)

    Masao, T.; Takasaki, Y.; Hirasawa, M.; Okajima, M.; Yamamoto, S.; Kawata, E.; Koori, Y.; Ochiai, S.; Shimizu, N.

    1980-01-01

    This paper is concerned with experimental and analytical studies to investigate dynamic behavior of deeply embedded structures such as nuclear reactor buildings. The principal points studied are as follows: (1) Examination of stiffness and radiation damping effects according to embedded depth, (2) verification for distributions of earth pressure according to embedded depth, (3) differences of response characteristics during oscillation according to embedded depth, and (4) proposal of an analytical method for seismic design. Experimental studies were performed by two ways: forced vibration test, and earthquake observation against a rigid body model embedded in soil. Three analytical procedures were performed to compare experimental results and to examine the relation between each procedure. Finally, the dynamic behavior for nuclear reactor buildings with different embedded depths were evaluated by an analytical method. (orig.)

  4. Sulphate in Liquid Nuclear Waste: from Production to Containment

    Energy Technology Data Exchange (ETDEWEB)

    Lenoir, M.; Grandjean, A.; Ledieu, A.; Dussossoy, J.L.; Cau Dit Coumes, C.; Barre, Y.; Tronche, E. [CEA Marcoule, DEN/DTCD/SECM/LDMC, Batiment 208 BP17171, Bagnols sur Ceze, 30207 (France)

    2009-06-15

    Nuclear industry produces a wide range of low and intermediate level liquid radioactive wastes which can include different radionuclides such as {sup 90}Sr. In La Hague reprocessing plant and in the nuclear research centers of CEA (Commissariat a l'Energie Atomique), the coprecipitation of strontium with barium sulphate is the technique used to treat selectively these contaminated streams with the best efficiency. After the decontamination process, low and intermediate level activity wastes incorporating significant quantities of sulphate are obtained. The challenge is to find a matrix easy to form and with a good chemical durability which is able to confine this kind of nuclear waste. The current process used to contain sulphate-rich nuclear wastes is bituminization. However, in order to improve properties of containment matrices and simplify the process, CEA has chosen to supervise researches on other materials such as cements or glasses. Indeed, cements are widely used for the immobilization of a variety of wastes (low and intermediate level wastes) and they may be an alternative matrix to bitumen. Even if Portland cement, which is extensively used in the nuclear industry, presents some disadvantages for the containment of sulphate-rich nuclear wastes (risk of swelling and cracking due to delayed ettringite formation), other cement systems, such as calcium sulfo-aluminate binders, may be valuable candidates. Another matrix to confine sulphate-rich waste could be the glass. One of the advantages of this material is that it could also immobilize sulphate containing high level nuclear waste which is present in some countries. This waste comes from the use of ferrous sulfamate as a reducing agent for the conversion of Pu{sup 4+} to Pu{sup 3+} in the partitioning stage of the actinides during reprocessing. Sulphate solubility in borosilicate glasses has already been studied in CEA at laboratory and pilot scales. At a pilot scale, low level liquid waste has been

  5. Dynamic analysis of auxiliary buildings in nuclear power plants

    International Nuclear Information System (INIS)

    Subramanian, K.V.; Madhava Rao, A.S.; Warudkar, A.S.

    1989-01-01

    All nuclear power plants have a large number of auxiliary buildings housing various services and control systems required for the operation of the plant. Illustrative examples are turbine building, control building, service building etc. These buildings are seismically qualified as Class I or Class II structures. Usually, these auxiliary buildings are of low rise type with two or three floors and floor heights varying from five to eight meters and of framed construction in steel or concrete or a combination of both the materials. The floors are usually staggered with large cutouts and may not extend over the full area in plan. Some of the bays are often of double story height with the columns continuous over a story in order to accommodate cranes and other equipment. The structural elements supporting the roof may consist of steel roof trusses instead of beams. The seismic analysis of these structures involves the formulation of the analytical model that can simulate the physical behavior of the structure as close as possible taking into consideration the practical aspects. The criteria adopted to formulate the mathematical model has an important bearing on the evaluated dynamic characteristics and seismic response

  6. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    Energy Technology Data Exchange (ETDEWEB)

    Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2011-03-15

    Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most

  7. Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

    International Nuclear Information System (INIS)

    Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.

    2011-01-01

    Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is

  8. Progressive building methods in nuclear power plant construction

    International Nuclear Information System (INIS)

    Sikura, V.

    1980-01-01

    A detailed description is given of the new prospective technologies used in the construction of the Bohunice V-1 nuclear power plant. They include the use of steel and large-area forms, the use of profile sheets as a substitute for forms, assembled raw partitions, wall shells consisting of clamped porous concrete slabs, assembled roof shells, special finish of concrete walls, bearing wall deep foundation, the use of modern building machinery for concreting, reinforcing and welding works. (M.S.)

  9. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    Dana, W.P.

    1995-12-01

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  10. Nuclear containment steel liner corrosion workshop : final summary and recommendation report.

    Energy Technology Data Exchange (ETDEWEB)

    Erler, Bryan A. (Erler Engineering Ltd., Chicago, IL); Weyers, Richard E. (Virginia Tech University, Blacksburg, VA); Sagues, Alberto (University of South Florida, Tampa, FL); Petti, Jason P.; Berke, Neal Steven (Tourney Consulting Group, LLC, Kalamazoo, MI); Naus, Dan J. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2011-07-01

    This report documents the proceedings of an expert panel workshop conducted to evaluate the mechanisms of corrosion for the steel liner in nuclear containment buildings. The U.S. Nuclear Regulatory Commission (NRC) sponsored this work which was conducted by Sandia National Laboratories. A workshop was conducted at the NRC Headquarters in Rockville, Maryland on September 2 and 3, 2010. Due to the safety function performed by the liner, the expert panel was assembled in order to address the full range of issues that may contribute to liner corrosion. This report is focused on corrosion that initiates from the outer surface of the liner, the surface that is in contact with the concrete containment building wall. Liner corrosion initiating on the outer diameter (OD) surface has been identified at several nuclear power plants, always associated with foreign material left embedded in the concrete. The potential contributing factors to liner corrosion were broken into five areas for discussion during the workshop. Those include nuclear power plant design and operation, corrosion of steel in contact with concrete, concrete aging and degradation, concrete/steel non-destructive examination (NDE), and concrete repair and corrosion mitigation. This report also includes the expert panel member's recommendations for future research.

  11. Role of Nuclear Knowledge Infrastructure in Building and Developing a Nuclear Industry

    International Nuclear Information System (INIS)

    Pershukov, V.

    2016-01-01

    Full text: Current level of development of nuclear technology pushes future owners and users to address several important challenges at the very beginning of preparation for introduction of nuclear power programme including development of national expertise and technical competence for safe and secure use of nuclear technology and creation of national scientific and research infrastructure. However, many times national stakeholders involved into the development of nuclear power programmes act in separate information environments. This creates confusion and may result in delays of execution of a nuclear power programmes. A common environment in this case is important to ensure information exchange and collaborative work on nuclear power programme development. A common environment facilitates transfer, preservation and spread of nuclear knowledge to all stakeholders involved in a national nuclear power programme. Due to the long life cycle of nuclear power plants, strengthening and maintaining the effective management of knowledge and information over the entire life cycle for licensed nuclear facilities is imperative. This covers areas including design, construction, commissioning, operation and decommissioning, especially in newcomer countries and new builds. Nuclear knowledge infrastructure assumes this role and helps countries deciding to develop nuclear programmes ensure the safe and secure use of its nuclear power for national development. (author

  12. International nuclear fuel cycle fact book. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  13. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

  14. Structural design and dynamic analysis of underground nuclear reactor containments

    International Nuclear Information System (INIS)

    Kierans, T.W.; Reddy, D.V.; Heale, D.G.

    1975-01-01

    Present actual experience in the structural design of undeground containments is limited to only four rather small reactors all located in Europe. Thus proposals for future underground reactors depend on the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. An application of such considerations in a recent input for the Underground Containment sub-section of the Seismic Task Group Report to the ASCE Committee for Nuclear Structures and Materials is presented as follows: underground concept considerations, siting criteria and structural selection, structural types, analytical and semi-analytical approaches, design and other miscellaneous considerations

  15. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  16. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    International Nuclear Information System (INIS)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

    1997-01-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments

  17. Study on alternatives of inertisation of nuclear power plant containment

    International Nuclear Information System (INIS)

    Baron, J.H.; Zarate, S.M.

    1998-01-01

    In the course of a severe accident in a nuclear power plant, the hydrogen generation and other flammable gases, during the core degradation phase and the interaction corium-concrete, could produce the failure of the containment by overpressure of by combustion. According to the analysis of the potential effects of hydrogen evolution, following accidents inside the containment trough a Defense-in depth principle, which attempts to assure that the containment must not fail catastrophically, two techniques have been evaluated: a: Inertisation pre-accident and b: Inertisation post-accident. The technique of inertisation pre-accident consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ) during the normal operation. The inertisation post-accident in combination with early venting system consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ), immediately after the beginning of the accident, while the radioactivity is still negligible inside the containment. A system of inertisation pre-accident with nitrogen is used on BWR Mark I and Mark II. Investigations on the inertisation post-accident of the containment atmosphere during severe accidents have been carried out with different objectives from principles of the decade of 1980. Studies concerning hydrogen problem for the nuclear power plants Atucha I and CAREM-25 have permitted to know that the hydrogen generation during an accidental sequence with core degradation, would result important, being able to arrive to form explosive mixtures. In the present work, the applicability of the techniques of inertisation is analyzed for the containment of the Atucha I and CAREM-25, considering the particular design characteristics of these plants. (author) [es

  18. Joint Arab project for building of nuclear power plants

    International Nuclear Information System (INIS)

    Mosbah, D.S.

    2010-01-01

    Recently many Arab countries have expressed their interest in adopting nuclear power for electricity generation and seawater desalination in their energy strategies and hence sought assistance from IAEA and Arab Atomic Energy Agency. The Council of the League of Arab States at the summit level encouraged member states to develop peaceful use of atomic energy in different aspects of development, especially in energy generation and to establish a cooperative Arab program in this field. The burden of infrastructure can be reduced significantly if an Arab country forms a sharing partnership with other Arab countries. The sharing can be between two or more Arab states. It can include physical facilities, common programmes and knowledge, which will reflect in economic benefits. The sharing can also contribute in a significant manner to harmonization of codes and standards in general and regulatory framework in particular. This article outlines the major direct benefits of joint nuclear Arab programme to build a Nuclear Power Plants. (author)

  19. Feasibility study of a contained pulsed nuclear propulsion engine

    International Nuclear Information System (INIS)

    Parlos, A.G.; Metzger, J.D.

    1994-01-01

    The result of a feasibility analysis of a contained pulsed nuclear propulsion (CPNP) engine concept utilizing the enormously dense energy generated by small nuclear detonations is presented in this article. This concept was initially proposed and studied in the 1950s and 1960s under the program name HELIOS. The current feasibility of the concept is based upon materials technology that has advanced to a state that allows the design of pressure vessels required to contain the blast associated with small nuclear detonations. The impulsive nature of the energy source provides the means for circumventing the materials thermal barriers that are inherent in steady-state nuclear propulsion concepts. The rapid energy transfer to the propellant results in high thrust levels for times less than 1 s following the detonation. The preliminary feasibility analysis using off-the-shelf materials technology appears to indicate that the CPNP concept can have thrust-to-weight ratios on the order of 1 or greater. Though the specific impulse is not a good indicator for impulsive engines, an operating-cycle averaged specific impulse of approximately 1000 or greater seconds was calculated. 16 refs

  20. Differential lead retention in zircons: implications for nuclear waste containment.

    Science.gov (United States)

    Gentry, R V; Sworski, T J; McKown, H S; Smith, D H; Eby, R E; Christie, W H

    1982-04-16

    An innovative ultrasensitive technique was used for lead isotopic analysis of individual zircons extracted from granite core samples at depths of 960, 2170, 2900, 3930, and 4310 meters. The results show that lead, a relatively mobile element compared to the nuclear waste-related actinides uranium and thorium, has been highly retained at elevated temperatures (105 degrees to 313 degrees C) under conditions relevant to the burial of synthetic rock waste containers in deep granite holes.

  1. Nuclear resonance vibrational spectroscopic studies of iron-containing biomolecules

    International Nuclear Information System (INIS)

    Ohta, Takehiro; Seto, Makoto

    2014-01-01

    In this review, we report recent nuclear resonance vibrational spectroscopic (NRVS) studies of iron-containing biomolecules and their model complexes. The NRVS is synchrotron-based element-specific vibrational spectroscopic methods. Unlike Raman and infrared spectroscopy, the NRVS can investigate all iron motions without selection rules, which provide atomic level insights into the structure/reactivity correlation of biologically relevant iron complexes. (author)

  2. Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

    International Nuclear Information System (INIS)

    Hirasawa, Yoshiya; Kigoshi, Yasutane; Yonenaga, Haruo.

    1992-01-01

    A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

  3. Influence of Steel Fibers on the Structural Performance of a Prestressed Concrete Containment Building

    International Nuclear Information System (INIS)

    Choun, Youngsun; Hahm, Daegi; Park, Junhee

    2013-01-01

    A large number of previous experimental investigations indicate that the use of steel fibers in conventional reinforced concrete (RC) can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity, and a high shear resistance under cyclic loadings will increase the seismic resisting capacity. In this study, the effects of steel fiber reinforced concrete (SFRC) on the ultimate pressure and seismic capacities of a PCCB are investigated. The effects of steel fibers on the ultimate pressure and shear resisting capacities of a PCCB are investigated. It is revealed that both of the ultimate pressure capacity and the shear resisting capacity of a PCCB can be greatly enhanced by introducing steel fibers in a conventional RC. Estimation results indicate that the ultimate pressure capacity and maximum lateral displacement of a PCCB can be improved by 16% and 64%, respectively, if a conventional RC contains hooked steel fibers in a volume fraction of 1.0%

  4. Thermal stress analysis of reactor containment building considering severe weather condition

    International Nuclear Information System (INIS)

    Lee, Yun; Kim, Yun-Yong; Hyun, Jung-Hwan; Kim, Do-Gyeum

    2014-01-01

    Highlights: • We examine that through-wall crack risk in cold weather is high. • It is predicted that cracking in concrete wall will not happen in hot region. • Cracking due to hydration heat can be controlled by appropriate curing condition. • Temperature differences between inner and outer face is relatively small in hot weather. - Abstract: Prediction of concrete cracking due to hydration heat in mass concrete such as reactor containment building (RCB) in nuclear power plant is a crucial issue in construction site. In this study, the numerical analysis for heat transfer and stress development is performed for the containment wall in RCB by considering the severe weather conditions. Finally, concrete cracking risk in hot and cold weather is discussed based on analysis results. In analyses considering severe weather conditions, it is found that the through-wall cracking risk in cold weather is high due to the abrupt temperature difference between inside concrete and the ambient air in cold region. In hot weather, temperature differences between inner and outer face is relatively small, and accordingly the relevant cracking risk is relatively low in contrast with cold weather

  5. Influence of Steel Fibers on the Structural Performance of a Prestressed Concrete Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Hahm, Daegi; Park, Junhee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    A large number of previous experimental investigations indicate that the use of steel fibers in conventional reinforced concrete (RC) can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity, and a high shear resistance under cyclic loadings will increase the seismic resisting capacity. In this study, the effects of steel fiber reinforced concrete (SFRC) on the ultimate pressure and seismic capacities of a PCCB are investigated. The effects of steel fibers on the ultimate pressure and shear resisting capacities of a PCCB are investigated. It is revealed that both of the ultimate pressure capacity and the shear resisting capacity of a PCCB can be greatly enhanced by introducing steel fibers in a conventional RC. Estimation results indicate that the ultimate pressure capacity and maximum lateral displacement of a PCCB can be improved by 16% and 64%, respectively, if a conventional RC contains hooked steel fibers in a volume fraction of 1.0%.

  6. First-passage Probability Estimation of an Earthquake Response of Seismically Isolated Containment Buildings

    International Nuclear Information System (INIS)

    Hahm, Dae-Gi; Park, Kwan-Soon; Koh, Hyun-Moo

    2008-01-01

    The awareness of a seismic hazard and risk is being increased rapidly according to the frequent occurrences of the huge earthquakes such as the 2008 Sichuan earthquake which caused about 70,000 confirmed casualties and a 20 billion U.S. dollars economic loss. Since an earthquake load contains various uncertainties naturally, the safety of a structural system under an earthquake excitation has been assessed by probabilistic approaches. In many structural applications for a probabilistic safety assessment, it is often regarded that the failure of a system will occur when the response of the structure firstly crosses the limit barrier within a specified interval of time. The determination of such a failure probability is usually called the 'first-passage problem' and has been extensively studied during the last few decades. However, especially for the structures which show a significant nonlinear dynamic behavior, an effective and accurate method for the estimation of such a failure probability is not fully established yet. In this study, we presented a new approach to evaluate the first-passage probability of an earthquake response of seismically isolated structures. The proposed method is applied to the seismic isolation system for the containment buildings of a nuclear power plant. From the numerical example, we verified that the proposed method shows accurate results with more efficient computational efforts compared to the conventional approaches

  7. Nuclear criticality safety calculational analysis for small-diameter containers

    International Nuclear Information System (INIS)

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

    1995-11-01

    This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant

  8. Structural review of the Palisades Nuclear Power Plant Unit 1 containment structure under combined loads for the Systematic Evaluation Program

    International Nuclear Information System (INIS)

    Liaw, C.Y.; Debeling, A.; Tsai, N.C.

    1981-12-01

    A structural reassessment of the containment structure of the Palisades Nuclear Power Plant Unit 1 was performed for the Nuclear Regulatory Commission as part of the Systematic Evaluation Program. Conclusions about the ability of the containment structure to withstand the Abnormal/Extreme Environment are presented. The reassessment focused mainly on the overall structural integrity of the containment building for the Abnormal/Extreme Environment. In this case, the Abnormal Environmental condition is caused by the worst case of either a Loss-of-Coolant Accident or a main steam line break. The Extreme Environmental condition is the Safe Shutdown Earthquake

  9. Nuclear containment structure subjected to commercial and fighter aircraft crash

    International Nuclear Information System (INIS)

    Sadique, M.R.; Iqbal, M.A.; Bhargava, P.

    2013-01-01

    Highlights: • Nuclear containment response has been studied against aircraft crash. • Concrete damaged plasticity and Johnson–Cook elasto-viscoplastic models were employed. • Boeing 747-400 and Boeing 767-400 aircrafts caused global failure of containment. • Airbus A320 and Boeing 707-320 aircrafts caused local damage. • Tension damage of concrete was found more prominent compared to compression damage. -- Abstract: The response of a boiling water reactor (BWR) nuclear containment vessel has been studied against commercial and fighter aircraft crash using a nonlinear finite element code ABAQUS. The aircrafts employed were Boeing 747-400, Boeing 767-400, Airbus A-320, Boeing 707-320 and Phantom F4. The containment was modeled as a three-dimensional deformable reinforced concrete structure while the loading of aircraft was assigned using the respective reaction–time curve. The location of strike was considered near the junction of dome and cylinder, and the angle of incidence, normal to the containment surface. The material behavior of the concrete was incorporated using the damaged plasticity model while that of the reinforcement, the Johnson–Cook elasto-viscoplastic model. The containment could not sustain the impact of Boeing 747-400 and Boeing 767-400 aircrafts and suffered rupture of concrete around the impact region leading to global failure. On the other hand, the maximum local deformation at the point of impact was found to be 0.998 m, 0.099 m, 0.092 m, 0.089 m, and 0.074 m against Boeing 747-400, Phantom F4, Boeing 767, Boeing 707-320 and Airbus A-320 aircrafts respectively. The results of the present study were compared with those of the previous analytical and numerical investigations with respect to the maximum deformation and overall behavior of the containment

  10. Nuclear containment structure subjected to commercial and fighter aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Sadique, M.R., E-mail: rehan.sadique@gmail.com; Iqbal, M.A., E-mail: iqbalfce@iitr.ernet.in; Bhargava, P., E-mail: bhpdpfce@iitr.ernet.in

    2013-07-15

    Highlights: • Nuclear containment response has been studied against aircraft crash. • Concrete damaged plasticity and Johnson–Cook elasto-viscoplastic models were employed. • Boeing 747-400 and Boeing 767-400 aircrafts caused global failure of containment. • Airbus A320 and Boeing 707-320 aircrafts caused local damage. • Tension damage of concrete was found more prominent compared to compression damage. -- Abstract: The response of a boiling water reactor (BWR) nuclear containment vessel has been studied against commercial and fighter aircraft crash using a nonlinear finite element code ABAQUS. The aircrafts employed were Boeing 747-400, Boeing 767-400, Airbus A-320, Boeing 707-320 and Phantom F4. The containment was modeled as a three-dimensional deformable reinforced concrete structure while the loading of aircraft was assigned using the respective reaction–time curve. The location of strike was considered near the junction of dome and cylinder, and the angle of incidence, normal to the containment surface. The material behavior of the concrete was incorporated using the damaged plasticity model while that of the reinforcement, the Johnson–Cook elasto-viscoplastic model. The containment could not sustain the impact of Boeing 747-400 and Boeing 767-400 aircrafts and suffered rupture of concrete around the impact region leading to global failure. On the other hand, the maximum local deformation at the point of impact was found to be 0.998 m, 0.099 m, 0.092 m, 0.089 m, and 0.074 m against Boeing 747-400, Phantom F4, Boeing 767, Boeing 707-320 and Airbus A-320 aircrafts respectively. The results of the present study were compared with those of the previous analytical and numerical investigations with respect to the maximum deformation and overall behavior of the containment.

  11. Building tomorrow's nuclear power plants with 4+D VR technology

    International Nuclear Information System (INIS)

    Lee, Il S.; Yoon, Sang H.; Shim, Kyu W.; Yu, Yong H.; Suh, Kune Y.

    2002-01-01

    There continues to be an increasing demand of electricity around the globe to fuel the industrial growth and to promote the human welfare. The economic activities have brought about richness in our material and cultural lives, in which process the electric power has been at the heart of the versatile energy sources. In order to timely and competitively respond to rapidly changing energy environment in the twenty-first century there is a growing need to build the advanced nuclear power plants in the unlimited workspace of virtual reality (VR) prior to commissioning. One can then realistically evaluate their construction time and cost per varying methods and options available from the leading-edge technology. In particular a great deal of efforts have yet to be made for time- and cost-dependent plant simulation and dynamically coupled database construction in the VR space. The operator training and personnel education may also benefit from the VR technology. The present work is being proposed in the three-dimensional space and time plus cost coordinates, i. e. four plus dimensional (4 + D) coordinates. The 4 + D VR application will enable the nuclear industry to narrow the technological gap from the other leading industries that have long since been employing the VR engineering. The 4 + D technology will help nurture public understanding of the special discipline of nuclear power plants. The technology will also facilitate public access to the knowledge on the nuclear science and engineering which has so far been monopolized by the academia, national laboratories and the heavy industry. The 4 + D virtual design and construction will open up the new horizon for revitalization of the nuclear industry over the globe in the foreseeable future. Considering the long construction and operation time for the nuclear power plants, the preliminary VR simulation capability for the plants will supply the vital information not only for the actual design and construction of the

  12. Containment analysis for the simultaneous detonation of two nuclear explosives

    International Nuclear Information System (INIS)

    Terhune, R.W.; Glenn, H.D.; Burton, D.E.; Rambo, J.T.

    1977-01-01

    The explosive phenomenology associated with the simultaneous detonation of two 2.2-kt nuclear explosives is examined. A comprehensive spatial-time pictorial of the resultant shock-wave phenomenology is given. The explosives were buried at depths of 200 m and 280 m, corresponding to a separation of approximately 4 final cavity radii. Constitutive relations for the surrounding medium were derived from the geophysical logs and core samples taken from an actual emplacement configuration at the Nevada Test Site. Past calculational studies indicate that successful containment may depend upon the development of a strong tangential-stress field (or ''containment cage'') surrounding the cavity at late times. A series of conditions that must be met to insure formation of this cage are presented. Calculational results, based on one- and two-dimensional finite-difference codes of continuum mechanics, describe how each condition has been fulfilled and illustrate the dynamic sequence of events important to the formation of the containment cage. They also indicate, at least for the geological site chosen, that two nuclear explosives do not combine to threaten containment

  13. Seismic analysis of liquid storage container in nuclear reactors

    International Nuclear Information System (INIS)

    Zhang Zhengming; He Shuyan; Xu Ming

    2007-01-01

    Seismic analysis of liquid storage containers is always difficult in the seismic design of nuclear reactor equipment. The main reason is that the liquid will generate significant seismic loads under earthquake. These dynamic liquid loads usually form the main source of the stresses in the container. For this kind of structure-fluid coupling problem, some simplified theoretical methods were usually used previously. But this cannot satisfy the requirements of engineering design. The Finite Element Method, which is now full developed and very useful for the structural analysis, is still not mature for the structure-fluid coupling problem. This paper introduces a method suitable for engineering mechanical analysis. Combining theoretical analysis of the dynamic liquid loads and finite element analysis of the structure together, this method can give practical solutions in the seismic design of liquid storage containers

  14. Building Nuclear Safety and Security Culture Within Regulatory Body

    International Nuclear Information System (INIS)

    Huda, K.

    2016-01-01

    To achieve a higher level of nuclear safety and security, it needs to develop the safety and security culture not only in the facility but also in the regulatory body. The regulatory body, especially needs to develop the safety and security culture within the organization, because it has a function to promote and oversee the culture in the facilities. In this sense, the regulatory body should become a role model. Development of the nuclear safety and security culture should be started by properly understanding its concept and awakening the awareness of individual and organization on the importance of nuclear safety and security. For effectiveness of the culture development in the regulatory body, the following steps are suggested to be taken: setting up of the regulatory requirements, self-assessment, independent assessment review, communication with the licensee, oversight of management system implementation, and integration with regulatory activities. The paper discusses those steps in the framework of development of nuclear safety and security culture in the regulatory body, as well as some important elements in building of the culture in the nuclear facilities. (author)

  15. Evaluation for rigidity of box construction of nuclear reactor building

    International Nuclear Information System (INIS)

    Yamakawa, Tetsuo

    1979-01-01

    A huge box-shaped structure (hereafter, called box construction) of reinforced concrete is presently utilized as the reactor building structure in nuclear power plants. Evaluation of the rigidity of the huge box construction is required for making a vibration analysis model of nuclear reactor buildings. It is necessary to handle the box construction as the plates to which the force in plane is applied. This paper describes that the bending theory in elementary beam theory is equivalent to a peculiar, orthogonally anisotropic plate, the shearing rigidity and film rigidity in y direction of which are put to infinity and the Poisson's ratio is put to zero, viewed from the two-dimensional theory of elasticity. The form factor of 1.2 for shearing deformation in rectangular cross section was calculated from the parabolic distribution of shearing stress intensity, and it is the maximum value. The factor is equal to 1.2 for slender beams, but smaller than 1.2 for short and thick beams, having tendency to converge to 1.0. The non-conformity of boundary conditions regarding the shearing force at the both ends of cantilevers does not affect very seriously the evaluation of shearing rigidity. From the above results, it was found that the application of the theory to the box construction was able to give the rigidity evaluation with sufficient engineering accuracy. The theory can also be applied to the evaluation of tube type ultrahigh buildings. (Wakatsuki, Y.)

  16. Photonuclear-based Detection of Nuclear Smuggling in Cargo Containers

    Science.gov (United States)

    Jones, J. L.; Haskell, K. J.; Hoggan, J. M.; Norman, D. R.; Yoon, W. Y.

    2003-08-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Los Alamos National Laboratory (LANL) have performed experiments in La Honda, California and at the Idaho Accelerator Center in Pocatello, Idaho to assess and develop a photonuclear-based detection system for shielded nuclear materials in cargo containers. The detection system, measuring photonuclear-related neutron emissions, is planned for integration with the ARACOR Eagle Cargo Container Inspection System (Sunnyvale, CA). The Eagle Inspection system uses a nominal 6-MeV electron accelerator and operates with safe radiation exposure limits to both container stowaways and to its operators. The INEEL has fabricated custom-built, helium-3-based, neutron detectors for this inspection application and is performing an experimental application assessment. Because the Eagle Inspection system could not be moved to LANL where special nuclear material was available, the response of the Eagle had to be determined indirectly so as to support the development and testing of the detection system. Experiments in California have successfully matched the delayed neutron emission performance of the ARACOR Eagle with that of the transportable INEEL electron accelerator (i.e., the Varitron) and are reported here. A demonstration test is planned at LANL using the Varitron and shielded special nuclear materials within a cargo container. Detector results are providing very useful information regarding the challenges of delayed neutron counting near the photofission threshold energy of 5.5 - 6.0 MeV, are identifying the possible utilization of prompt neutron emissions to allow enhanced signal-to-noise measurements, and are showing the overall benefits of using higher electron beam energies.

  17. Non destructive evaluation of containment nuclear plants structures

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, V. [Aix Marseille Univ., Aix en Provence (France). LMA, CNRS UPR 7051, IUT; Verdier, J. [Toulouse Univ. (France). UPS, INSA, LMDC; Sbartai, Z.M. [Bordeaux Univ., Talence (France). I2M; and others

    2015-07-01

    French Projects of Investment for the Future, called ''Research for Nuclear Safety and Radiation Protection'' have been initiated to further research on the causes, the management, the impact of the observed nuclear accidents and to propose and validate solutions to limit the risk and the consequences. In this context the ''Non Destructive Evaluation of nuclear plants containment'' project (ENDE) with eight partners (six research institutes and two industrials) supported by the ''National Agency of Research'', proposes a methodology for the Non Destructive Evaluation of the containment capacity to fulfil its two major functions: strength and leak tightness. The NDE measurements will be performed under conditions representing the specific solicitations of a decennial inspection, and after or during a reference accident. The project aims to develop NDEs, to combine them by data fusion and to select the most efficient combinations with quantitative criteria. The work is based on: - Structuring the knowledge and developing an experimental plan. - Evaluating the material in representative conditions of accidental solicitations (water saturation, porosity, strength, elastic modulus, stress) and the diffuse thermal damage (micro cracks) - Monitoring the transition from diffuse to continuous damage (cracks) and monitoring a crack under stress (opening and width). - Implementing ND Techniques on-site. The ND techniques retained after selection will be implemented on a containment mock-up on the 1/3 scale. This mock-up developed by EDF (Electricite de France) will be available in 2016. It will be comparable to those of real size containment regarding pressure and temperature conditions. The measures deduced from the NDEs will be introduced in another project (Macena) that aims to simulate the water and heat transfers as well as creep occurring in a reference accident. We will present the methodology and the results

  18. Assessment of materials for nuclear fuel immobilization containers

    Energy Technology Data Exchange (ETDEWEB)

    Nuttall, K; Urbanic, V F

    1981-09-01

    A wide range of engineering metals and alloys has been assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The expected range of service conditions in the disposal vault are discussed, as well as the material properties required for this application. An important requirement is that the container last at least 500 years without being breached. The assessment is treated in two parts. Part I concentrates on the physical and mechanical metallurgy, with special reference to strength, weldability, potential embrittlement mechanisms and some economic aspects. Part II discusses possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditions in the vault. Localized corrosion and delayed fracture processes are identified as being most likely to limit container lifetime. Hence an essential requirement is that such processes either be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further consideration as possible container materials: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the enviromental conditions in the vault. 42 figures, 31 tables.

  19. Lead corrosion evaluation in high activity nuclear waste container (Argentina)

    International Nuclear Information System (INIS)

    Guasp, R.; Lanzani, L.; Bruzzoni, P.; Cufre, W.; Semino, C.J.

    2000-01-01

    This report describes a study of high activity nuclear waste canister corrosion in a deep geological disposal. In this canister design, the vitrified nuclear waste stainless steel container is shielded by a 100 mm thick lead wall. For mechanical resistance, the canister will also have a thin carbon steel external liner. Experimental and mathematical modeling studies are aimed to asses the corrosion kinetics of the carbon steel liner in first instance and then, once this liner has been corroded away, the corrosion kinetics of the main lead barrier. Being that oxygen reduction is the main cathodic reaction that supports the anodic oxidation of iron, a model is described predicting the rate of oxygen consumption in a sealed deep nuclear waste disposal vault as a result of the canister corrosion. Oxidation processes other than container corrosion, and that can account also for oxygen depletion, are not taken into consideration. Corrosion experimental studies on lead and its alloys in groundwater are also reported. These experiments are aimed to improve the corrosion resistance of commercial lead in groundwater. (author)

  20. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  1. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  2. Effects of Containment on Radionuclide Releases from Underground Nuclear Explosions

    Science.gov (United States)

    Carrigan, C. R.; Sun, Y.

    2016-12-01

    Confirming the occurrence of an underground nuclear explosion can require capturing short-lived noble gas radioisotopes produced by the explosion, sometimes referred to as the "smoking gun" for nuclear explosion detection. It is well known that the radioisotopic distribution resulting from the detonation evolves with time in the explosion cavity. In effect, the explosion cavity or chimney behaves as a chemical reactor. As long as the parent and daughter radionuclides remain in a closed and well-mixed cavity, parameters, such as radioxenon isotopic ratios, can be calculated analytically from a decay-chain network model. When gases from the cavity migrate into the containment regime, consideration of a "leaky reactor" model is more appropriate. We consider several implications of such a leaky reactor model relevant to interpretations of gas samples from the subsurface during an on-site inspection that could potentially be carried out under the Comprehensive Nuclear Test Ban Treaty. Additionally, we have attempted to validate our leaky reactor model against atmospheric observations of radioactive xenon isotopes detected by radionuclide monitoring stations in Japan and Russia following the February 2013 DPRK underground nuclear explosion (Carrigan et al., 2016). While both model uncertainty and observational error are significant, our model of isotopic evolution appears to be in broad agreement with radionuclide observations, and for the first time links atmospheric measurements of radioxenon isotopic ratios to estimates of seismic yield. Carrigan et al., Scientific Reports 6, Article number: 23032 (2016) doi:10.1038/srep23032

  3. A study on the functional assessment of the prestressed system and main structural elements in life extended containment building

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, C. H.; Kim, S. W.; Choi, J. G. [DAEWOO E and C Institute of Costruction Technology, Suwon (Korea, Republic of)] (and others)

    2001-10-15

    The design life of KNGR (Korean Next Generation Reactor) containment buildings is extended from 40 years to 60 years. However, nuclear reactor buildings are passive structures that are impossible to be exchanged in the case of degradation by the deterioration and so on when extending the design life of structures. Therefore, it is necessary to consider the long-term safety endurance in the design and construction of KNGR. Also, ti is judged that choice of the material and various test methods should be prescribed clearly. In this study, the reduction schemes of deterioration and the safety-ensuring schemes are drawn for the expected performance to be maintained from the beginning of the service to the required period together with ensuring the safety and serviceability of KNGR which will be constructed with the design life of 60 years, taking into account the dimensions, selection of material and construction methods in the design and construction stages. Also, the validity is to be examined for the estimation method of long-term losses of stress introduced to KNGR whose design life is increased to 60 years. The durability enhancement scheme on the design and construction for the design life extension of nuclear containment buildings is to be drawn through these studies. These results are utilized as the basic data for the safety inspection and examination guides of KNGR and finally the additional investigations are proposed for the items which require long-term studies.

  4. New methods of building WWER 1000 nuclear power plants

    International Nuclear Information System (INIS)

    Tadzher, O.

    1983-01-01

    The shortening of construction time needed for building the 5th unit of the Kozloduj nuclear power plant in Bulgaria is discussed. In the 2nd to 4th stages of construction the builders are expected to use the large CC-4000 mobile crane with high lifting power (Demag, FRG) for construction and assembly work. This work may be done by ''open'' assembly, i.e., the assembly of equipment in an unfinished building. By using the said crane it is possible to increase the weight of the individual units and thereby to reduce their number. The advantages of ''open'' assembly is that there is no need for a opecial transport corridor for the lifting and assembly of heavy handling units. A preliminary evaluation of the new technology has shown that assembly of the reactor hall equipment can be started 6 to 8 months ahead of schedule. (E.S.)

  5. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    International Nuclear Information System (INIS)

    Noh, Hyung Gyun; Kang, Hie Chan

    2016-01-01

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

  6. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

    2016-05-15

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

  7. Alcohol-free alkoxide process for containing nuclear waste

    Science.gov (United States)

    Pope, James M.; Lahoda, Edward J.

    1984-01-01

    Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

  8. Detecting nuclear materials smuggling: performance evaluation of container inspection policies.

    Science.gov (United States)

    Gaukler, Gary M; Li, Chenhua; Ding, Yu; Chirayath, Sunil S

    2012-03-01

    In recent years, the United States, along with many other countries, has significantly increased its detection and defense mechanisms against terrorist attacks. A potential attack with a nuclear weapon, using nuclear materials smuggled into the country, has been identified as a particularly grave threat. The system for detecting illicit nuclear materials that is currently in place at U.S. ports of entry relies heavily on passive radiation detectors and a risk-scoring approach using the automated targeting system (ATS). In this article we analyze this existing inspection system and demonstrate its performance for several smuggling scenarios. We provide evidence that the current inspection system is inherently incapable of reliably detecting sophisticated smuggling attempts that use small quantities of well-shielded nuclear material. To counter the weaknesses of the current ATS-based inspection system, we propose two new inspection systems: the hardness control system (HCS) and the hybrid inspection system (HYB). The HCS uses radiography information to classify incoming containers based on their cargo content into "hard" or "soft" containers, which then go through different inspection treatment. The HYB combines the radiography information with the intelligence information from the ATS. We compare and contrast the relative performance of these two new inspection systems with the existing ATS-based system. Our studies indicate that the HCS and HYB policies outperform the ATS-based policy for a wide range of realistic smuggling scenarios. We also examine the impact of changes in adversary behavior on the new inspection systems and find that they effectively preclude strategic gaming behavior of the adversary. © 2011 Society for Risk Analysis.

  9. Supporting human performance in operations - principles for new nuclear build

    International Nuclear Information System (INIS)

    Lane, L.; Davey, E.

    2006-01-01

    Operational experience worldwide continues to demonstrate that human performance is a key factor in the ongoing safety, production, and protection of investment in operation of nuclear plants for electricity generation. Human performance in support of plant operational objectives can be influenced by a range of factors, for example: organizational culture and expectations; role assignments, training, and individual and team behaviours; and the support offered by the workplace environment, tools, and task design. This paper outlines a perspective on some of the principles that should be considered for application in the design of new nuclear build to facilitate support for human performance in plant operations. The principles identified focus on but are not limited to the tasks of shift staff, and are derived from the observations and experience of the authors who are experienced with control room operations in current plants. (author)

  10. Supporting human performance in operations - principles for new nuclear build

    Energy Technology Data Exchange (ETDEWEB)

    Lane, L. [Ontario Power Generation, Darlington Nuclear Div., Bowmanville, Ontario (Canada); Davey, E. [Crew Systems Solutions, Deep River, Ontario (Canada)

    2006-07-01

    Operational experience worldwide continues to demonstrate that human performance is a key factor in the ongoing safety, production, and protection of investment in operation of nuclear plants for electricity generation. Human performance in support of plant operational objectives can be influenced by a range of factors, for example: organizational culture and expectations; role assignments, training, and individual and team behaviours; and the support offered by the workplace environment, tools, and task design. This paper outlines a perspective on some of the principles that should be considered for application in the design of new nuclear build to facilitate support for human performance in plant operations. The principles identified focus on but are not limited to the tasks of shift staff, and are derived from the observations and experience of the authors who are experienced with control room operations in current plants. (author)

  11. Building a better future: Contributions of nuclear science and technology

    International Nuclear Information System (INIS)

    1998-02-01

    Food, health, energy, industry and the environment. The fields could not be more diverse. Yet they are all linked in one universe - the worldwide research and development activities of the IAEA's Department of Research and Isotopes. These are just a few of many examples where nuclear-based technologies are helping to understand and provide constructive solutions to today's economic and social problems. The development of these technologies through the IAEA's Research Contract Programme and their extension through its Technical Co-operation (TC) Programme and information services encourage specialists in Member States to transform scientific results into knowledge, products and services which benefit society at large. In each of these sectors, the Department's activities foster leadership and international cooperation in applying nuclear technology to everyday scientific, technological and socio-economic problems. Together with international meetings, they add to the collective knowledge within Member States that supports capacity building and end-user oriented development projects supported by the TC Programme

  12. An assessment of materials for nuclear fuel immobilization containers

    International Nuclear Information System (INIS)

    Nuttall, K.; Urbanic, V.F.

    1981-09-01

    A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

  13. Characterization of nuclear reactor containment penetrations. Preliminary report

    International Nuclear Information System (INIS)

    Bump, T.R.; Seidensticker, R.W.; Shackelford, M.A.; Gambhir, V.K.; McLennan, G.L.

    1984-06-01

    This report summarizes the survey work conducted by Argonne National Laboratory on the design and details of major penetrations in 22 nuclear power plants. The survey includes all containment types and materials in current use. It also includes details of all types of penetrations (except for electrical penetration assemblies and valves) and the seals and gaskets used in them. The report provides a test matrix for testing major penetrations and for testing seals and gaskets in order to evaluate their leakage potential under severe accident conditions

  14. Detection of nuclear material by photon activation inside cargo containers

    Science.gov (United States)

    Gmar, Mehdi; Berthoumieux, Eric; Boyer, Sébastien; Carrel, Frédérick; Doré, Diane; Giacri, Marie-Laure; Lainé, Frédéric; Poumarède, Bénédicte; Ridikas, Danas; Van Lauwe, Aymeric

    2006-05-01

    Photons with energies above 6 MeV can be used to detect small amounts of nuclear material inside large cargo containers. The method consists in using an intense beam of high-energy photons (bremsstrahlung radiation) in order to induce reactions of photofission on actinides. The measurement of delayed neutrons and delayed gammas emitted by fission products brings specific information on localization and quantification of the nuclear material. A simultaneous measurement of both of these delayed signals can overcome some important limitations due to matrix effects like heavy shielding and/or the presence of light elements as hydrogen. We have a long experience in the field of nuclear waste package characterization by photon interrogation and we have demonstrated that presently the detection limit can be less than one gram of actinide per ton of package. Recently we tried to extend our knowledge to assess the performance of this method for the detection of special nuclear materials in sea and air freights. This paper presents our first results based on experimental measurements carried out in the SAPHIR facility, which houses a linear electron accelerator with the energy range from 15 MeV to 30 MeV. Our experiments were also modeled using the full scale Monte Carlo techniques. In addition, and in a more general frame, due to the lack of consistent data on photonuclear reactions, we have been working on the development of a new photonuclear activation file (PAF), which includes cross sections for more than 600 isotopes including photofission fragment distributions and delayed neutron tables for actinides. Therefore, this work includes also some experimental results obtained at the ELSA electron accelerator, which is more adapted for precise basic nuclear data measurements.

  15. Axisymmetric analysis of a 1:6-scale reinforced concrete containment building using a distributed cracking model for the concrete

    International Nuclear Information System (INIS)

    Weatherby, J.R.

    1987-09-01

    Results of axisymmetric structural analyses of a 1:6 scale model of a reinforced concrete nuclear containment building are presented. Both a finite element shell analysis and a simplified membrane analysis were made to predict the structural response and ultimate pressure capacity of the model. Analytical results indicate that the model will fail at an internal pressure of 187 psig when the stress level in the hoop reinforcement at the midsection of the cylinder exceeds the ultimate strength of the bar splices. 5 refs., 34 figs., 6 tabs

  16. Study of the sodium oxide aerosol behavior in a containment building

    International Nuclear Information System (INIS)

    Fermandjian, Jean.

    1982-01-01

    Fast breeder reactor safety analysis needs validated models to predict the behavior of sodium oxide aerosols (released during hypothetical accidents) in the different reactor compartments and, in particular, to evaluate the quantity (and the size distribution) of the particles which can be released outside the containment building by taking into account the associated devices (ventilation, especially). In order to validate the computer models, experimental tests were performed by CEA/DSN (Atomic Energy Commission/Nuclear Safety Department) at Cadarache: sodium pool fire tests in a 400 m 3 concrete vessel (CASSANDRE tests) and in a 4.4 m 3 steel vessel (EMIS tests). Furthermore, we utilized the results of the experimental tests performed by our German colleagues of KFK/LAF (Kernforschungszentrum Karlsruhe/Laboratorium fuer Aerosolphysik und Filtertechnik): sodium pool fire tests in a 220 m 3 steel vessel (FAUNA tests). The application of the computer models to the experimental tests and to the reactor case allowed us: - to explain the differences observed between the results given by the codes ''with log-normal hypothesis'' (HAARM 3) and ''with finite difference'' (PARDISEKO 3B), - to define the limits of application of the codes ''with log-normal hypothesis'', - to determine the important parameters (characteristics of the aerosol source, form factors of the agglomerates and collision efficiency) and the important phenomena (turbulent coagulation, in particular) [fr

  17. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a 'limited propagation' argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J ox ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NANO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J ox are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained. (author)

  18. The stress corrosion cracking of copper nuclear waste containers

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.; Ikeda, B.M.

    1999-01-01

    The extent of stress corrosion cracking (SCC) of copper nuclear waste containers is being predicted on the basis of a limited propagation argument. In this argument, it is accepted that crack initiation may occur, but it is argued that the environmental conditions and material properties required for a through-wall crack to propagate will not be present. In this paper, the effect of one environmental parameter, the supply of oxidant (J OX ), on the crack growth rate is examined. Experiments have been conducted on two grades of Cu in NaNO 2 environments using two loading techniques. The supply of oxidant has been varied either electrochemically in bulk solution using different applied current densities or by embedding the loaded test specimens in compacted buffer material containing O 2 as the oxidant. Measured and theoretical crack growth rates as a function of J OX are compared with the predicted oxidant flux to the containers in a disposal vault and an estimate of the maximum crack depth on a container obtained

  19. Aging of concrete containment structures in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Mori, Yasuhiro; Arndt, E.G.

    1992-01-01

    Concrete structures play a vital role in the safe operation of all light-water reactor plants in the US Pertinent concrete structures are described in terms of their importance design, considerations, and materials of construction. Degradation factors which can potentially impact the ability of these structures to meet their functional and performance requirements are identified. Current inservice inspection requirements for concrete containments are summarized. A review of the performance history of the concrete components in nuclear power plants is provided. A summary is presented. A summary is presented of the Structural Aging (SAG) Program being conducted at the Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved bases for their continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technologies, and quantitiative methodology for continued service conditions. Objectives and a summary of accomplishments under each of these tasks are presented

  20. Physical modeling of spent-nuclear-fuel container

    Directory of Open Access Journals (Sweden)

    Wang Liping

    2012-11-01

    Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

  1. A feasibility study for a contained pulsed nuclear propulsion concept

    International Nuclear Information System (INIS)

    Parlos, A.G.; Metzger, J.D.

    1993-01-01

    A preliminary analysis of a pulsed propulsion concept is performed utilizing the enormously dense energy generated by small nuclear detonations. The concept feasibility is based on the premise that current materials technology has undergone significant breakthroughs, allowing design of pressure vessels capable of containing the blast associated with such detonations. Furthermore, the rapid energy transfer to the propellant, allows generation of high thrust levels for up to 10 ms following the detonation. Preliminary reevaluation of the concept using off-the-shelf materials technology appears to indicate that the contained pulsed nuclear propulsion concept has no major flaws, and it can provide thrust levels resulting in average thrust-to-weight ratios on the order of 2--2.5 over an engine operating cycle. Furthermore, even though the specific impulse is not a good performance indicator for impulsive engines, operating-cycle-averaged specific impulse of approximately 1800 s has been calculated. The engine mass associated with this performance is on the order of 50 Mg. The concept appears attractive for a number of missions planned for the Space Exploration Initiative, however, there are still a number of issues that must be addressed

  2. Alternative technology of containment construction for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Chalus, Z.

    1982-01-01

    A number of alternatives was assessed for the assembly of the steel elements of the cylindrical part of containment. Alternative 1 is based on the common technology of manufacture, transport and assembly of reinforced concrete blocks of ca. 3x12 m in size, used for building leak-proof walls of WWER 440 nuclear power plants. Alternative 2 is based on reinforced concrete blocks using 12x12 m blocks assembled from individual elements on the building. site. Alternative 3 is a specific variant of the previous alternative. Alternative 4 envisages the assembly of a prefabricated support structure made of steel. Alternative 5 is based on the gradual assembly of partial elements mounted onto a support structure. Alternative 6 only differs from 5 in the method of assembly and manufacture of the support structure. All alternatives are shown in diagrams. (J.B.)

  3. The Flooding Water Source Analysis following the Feed Line Break at the Compartment outside Containment for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Park, Young Chan [ACT, Daejeon (Korea, Republic of)

    2007-07-01

    The Periodic Safety Review(PSR) has been performing for the operating nuclear power plant in Korea. One of the PSR evaluation items is environmental qualification. Flooding issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feed water line region of intermediate building and lower floor. This study presents to analyze flood level of feed water line breaks for the Westinghouse nuclear power plant. This analyses provides the mass and energy releases using the developed methodology for a break outside containment. For the analyses RETRAN-3D computer program is used.

  4. Evaluation of the pressure loads generated by hydrogen explosion in auxiliary nuclear building

    International Nuclear Information System (INIS)

    Ahmed Bentaib; Alexandre Bleyer; Pierre Pailhories; Jean-Pierre L'heriteau; Bernard Chaumont; Jerome Dupas; Jerome Riviere

    2005-01-01

    Full text of publication follows: In the framework of nuclear safety, a hydrogen leaks in the auxiliary nuclear building would raise a explosion hazard. A local ignition of the combustible mixture would give birth initially to a slow flame, rapidly accelerated by obstacles. This flame acceleration is responsible for high pressure loads that can damage the auxiliary building and destroy safety equipments in it. In this paper, we evaluate the pressure loads generated by an hydrogen explosion for both bounding and realistic explosion scenarios. The bounding scenarios use stoichiometric hydrogen-air mixtures and the realistic scenarios correspond to hydrogen leaks with mass flow rate varying between 1 g/s and 9 g/s. For every scenario, the impact of the ignition location and ignition time are investigated. The hydrogen dispersion and explosion are computed using the TONUS code. The dispersion model used is based on a finite element solver and the explosion is simulated by a structured finite volumes EULER equation solver and the combustion model CREBCOM which simulates the hydrogen/air turbulent flame propagation, taking into account 3D complex geometry and reactants concentration gradients. The pressure loads computed are then used to investigate the occurrence of a mechanical failure of the tanks located in the auxiliary nuclear building and containing radioactive fluids. The EUROPLEXUS code is used to perform 3D mechanical calculations because the loads are non uniform and of rather short deviation. (authors)

  5. Nuclear shielding of openings in ITER Tokamak building

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

  6. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  7. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    International Nuclear Information System (INIS)

    Horschel, D.S.

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission's program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix

  8. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

  9. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  10. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan's effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  11. Method for the chemical reprocessing of irradiated nuclear fuels, in particular nuclear fuels containing uranium

    International Nuclear Information System (INIS)

    Koch, G.

    1976-01-01

    In the chemical processing of irradiated uranium-containing nuclear fuels which are hydrolyzed with aqueous nitric acid, a suggestion is made to use as quaternary ammonium nitrate trialkyl-methyl ammonium nitrates as extracting agent, in which the sum of C atoms is greater than 16. In the illustrated examples, tricaprylmethylammonium nitrate, trilaurylmethylammonium nitrate and tridecylmethylammonium nitrate are named. (HPH/LH) [de

  12. Evaluation of prestress losses in nuclear reactor containments

    International Nuclear Information System (INIS)

    Lundqvist, Peter; Nilsson, Lars-Olof

    2011-01-01

    Research highlights: → Prestress losses in reactor containments were estimated using prediction models. → The predicted prestress losses were compared to long-term measurements. → The accuracy of the models was improved by considering actual drying conditions. → Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage

  13. Evaluation of prestress losses in nuclear reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist, Peter, E-mail: peter.lundqvist@kstr.lth.s [Div. of Structural Engineering, Lund University, Lund (Sweden); Nilsson, Lars-Olof [Div. of Building Materials, Lund University, Lund (Sweden)

    2011-01-15

    Research highlights: Prestress losses in reactor containments were estimated using prediction models. The predicted prestress losses were compared to long-term measurements. The accuracy of the models was improved by considering actual drying conditions. Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage of concrete and

  14. Containment hydrogen removal system for a nuclear power plant

    International Nuclear Information System (INIS)

    Callaghan, V.M.; Flynn, E.P.; Pokora, B.M.

    1984-01-01

    A hydrogen removal system (10) separates hydrogen from the containment atmosphere of a nuclear power plant using a hydrogen permeable membrane separator (30). Water vapor is removed by condenser (14) from a gas stream withdrawn from the containment atmosphere. The gas stream is then compressed by compressor (24) and cooled (28,34) to the operating temperature of the hydrogen permeable membrane separator (30). The separator (30) separates the gas stream into a first stream, rich in hydrogen permeate, and a second stream that is hydrogen depleted. The separated hydrogen is passed through a charcoal adsorber (48) to adsorb radioactive particles that have passed through the hydrogen permeable membrane (44). The hydrogen is then flared in gas burner (52) with atmospheric air and the combustion products vented to the plant vent. The hydrogen depleted stream is returned to containment through a regenerative heat exchanger (28) and expander (60). Energy is extracted from the expander (60) to drive the compressor (24) thereby reducing the energy input necessary to drive the compressor (24) and thus reducing the hydrogen removal system (10) power requirements

  15. Prediction of prestressing losses for long term operation of nuclear reactor buildings

    Directory of Open Access Journals (Sweden)

    Thillard G.

    2011-04-01

    Full Text Available Prestressed concrete is used in nuclear reactor buildings to guarantee containment and structural integrity in case of an accident. Monitoring and operating experience over 40 years has shown that prestressing losses can be much greater than the design estimation based on the usual standard laws. A method was developed to determine the realistic residual prestress level in structures, in particular for those where no embedded instrumentation was installed, taking into account in situ measurement results rather than design characteristics. The results can enable the owner to justify extending the lifespan while guaranteeing adequate safety and to define and plan adequate maintenance actions.

  16. Nuclear emergency buildings of Asco and Vandellos II nuclear power plants; Centros alternativos de emergencias de las centrales nucleares de Asco y Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    Massuet, J.; Sabater, J.; Mirallas Esteban, S.

    2016-08-01

    The Nuclear Emergency Buildings sited at Asco and Vandellos II Nuclear Power Plants (NPP) are designed to safety manage emergencies in extreme situations, beyond the design basis of the Nuclear Power Plants. Designed in accordance with the requirements of the Spanish Nuclear Regulator (Consejo de Seguridad Nuclear-CSN) these buildings are ready to operate over a period of 72 hours without external assistance and ensure habitability for crews of 120 and 70 people respectively. This article describes the architectural conception, features and major systems of the Nuclear Emergency Buildings sited at Asco and Vandellos II. (Author)

  17. Development of finite element models for the study of ageing effects in CANDU 6 concrete containment buildings

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Y.; Jaffer, S., E-mail: Yuqing.Ding@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    In nuclear power plants (NPPs), concrete containment buildings (CCBs) provide the final physical barrier against the release of radioactive materials into the environment and protect the nuclear structures housed within the containment building. CCBs have to be maintained to ensure leak tightness and sound structural integrity for the safe operation throughout the life of NPPs. However, the integrity of CCBs may be affected by the ageing of its concrete, post-tensioning cables and reinforcing bars (rebars). Finite element models (FEMs) of CANDU 6 CCBs have been developed using 2 independent finite element programs for the study of the effect of ageing of CCBs. These FEMs have been validated using multiple-source data and have been used for preliminary analyses of the effect of thermal load and ageing degradation on the concrete structure. The modelling assumptions and simplifications, approach, and validation are discussed in this paper. The preliminary analyses for temperature effects and potential applications to the study of ageing degradation in CCBs using the FEMs are briefly introduced. (author)

  18. The build-up and characterization of nuclear burn-up wave in a fast ...

    Indian Academy of Sciences (India)

    K V Anoop

    2018-02-07

    Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

  19. A program to assess microbial impacts on nuclear waste containment

    International Nuclear Information System (INIS)

    Horn, J.; Meike, A.

    1996-01-01

    In this paper we discuss aspects of a comprehensive program to identify and bound potential effects of microorganisms on long-term nuclear waste containment, using as examples, studies conducted within the Yucca Mountain Project. A comprehensive program has been formulated which cuts across standard disciplinary lines to address the specific concerns of microbial activity in a radioactive waste repository. Collectively, this program provides bounding parameters of microbial activities that modify the ambient geochemistry and hydrology, modify corrosion rates, and transport and transform radionuclides under conditions expected to be encountered after geological waste emplacement. This program is intended to provide microbial reaction rates and bounding conditions in a form that can be integrated into existing chemical and hydrological models. The inclusion of microbial effects will allow those models to more accurately assess long term repository integrity

  20. Gamma radiation induced changes in nuclear waste glass containing Eu

    Science.gov (United States)

    Mohapatra, M.; Kadam, R. M.; Mishra, R. K.; Kaushik, C. P.; Tomar, B. S.; Godbole, S. V.

    2011-10-01

    Gamma radiation induced changes were investigated in sodium-barium borosilicate glasses containing Eu. The glass composition was similar to that of nuclear waste glasses used for vitrifying Trombay research reactor nuclear waste at Bhabha Atomic Research Centre, India. Photoluminescence (PL) and electron paramagnetic resonance (EPR) techniques were used to study the speciation of the rare earth (RE) ion in the matrix before and after gamma irradiation. Judd-Ofelt ( J- O) analyses of the emission spectra were done before and after irradiation. The spin counting technique was employed to quantify the number of defect centres formed in the glass at the highest gamma dose studied. PL data suggested the stabilisation of the trivalent RE ion in the borosilicate glass matrix both before and after irradiation. It was also observed that, the RE ion distributes itself in two different environments in the irradiated glass. From the EPR data it was observed that, boron oxygen hole centre based radicals are the predominant defect centres produced in the glass after irradiation along with small amount of E’ centres. From the spin counting studies the concentration of defect centres in the glass was calculated to be 350 ppm at 900 kGy. This indicated the fact that bulk of the glass remained unaffected after gamma irradiation up to 900 kGy.

  1. Earthquake response of nuclear reactor building deeply embedded in soil

    International Nuclear Information System (INIS)

    Masao, T.; Hirasawa, M.; Yamamoto, S.; Koori, Y.

    1977-01-01

    Regarding the earthquake response of nuclear reactor building embedded in soil, experimental and theoretical investigations has been performed on a model of height-3.75 meter, bottom cross section-5x5 meter, weight-173 ton made of conrete under the financial support of Japanese government (The Science and Technology Agency). The top of model was excited by an eccentric mass vibration that can generate maximum force of 3 tons. Earthpressures were measured at the bottom and side wall of model, and displacements were also measured at the top of model (6 components) and ground surface changed in the steps which were 0, 20, 47, 73, 100% (against the height of model). Using these experimental results and soil properties, dynamical characteristics were studied, including resonant frequency, radiation damping, vibrational mode, frequency response and earthpressure distribution around the model at varying embedment by lumped model, cyclindrical elastic wave model and FEM models (thin layer elements). (Auth.)

  2. A comprehensive framework for successful Nuclear New Build delivery

    International Nuclear Information System (INIS)

    Jimenez, Rafael J.

    2014-01-01

    This paper first addresses the concept of success, the factors which favours it and the best way in which these critical success factors should be embedded in the whole process of developing a new nuclear power plant (NPP). Taking as basic references the accumulated experience and lessons learnt from the past, the most recent developments and the relevant life-cycle criteria to measure the success of a project, the paper pass review to the different phases and stages of the nuclear power plant, highlighting the main areas of work and issues to be addressed, from the early days of its conception to its final closing and decommissioning. The purpose is not to enter into a high-level of detail in any particular aspect of project but to offer a quick 'all-in-one-glance' look at all the essential building-blocks and key drivers in the delivery of a new nuclear project, making sure that the framework touches on all the key concepts, check-lists, areas to work on, tools and issues to be addressed and presents them together as part of a comprehensive and cohesive picture. The aim would be to offer a very practical guide that did not miss any relevant part. Once the project (or programme) team gets a good grasp of the framework, and depending of the starting point and particular challenges for each project, the model will demand more or less focus in certain topics and it will be just a question of entering into as much level of detailed and granularity as needed. Developing a nuclear new build project is always a very ambitious enterprise. It is a venture we have to be humble about, thinking, planning and acting with maximum rigour, cautiousness and professionalism, not underestimating its 'grey zones' and certainly not overestimating our capabilities. It is a long term challenging journey, not precisely exempt of risks, obstacles and uncertainties, but a very thrilling and rewarding one worth the effort. Its successful outcome translates into many and

  3. Constitutive model for evaluation of nuclear containment structures

    Energy Technology Data Exchange (ETDEWEB)

    Gocevski, Vladimir [Hydro-Quebec, 75 Rene-Levesque Boulevard, West Montreal, QC H2Z 1A4 (Canada)

    2006-09-15

    The paper presents the new constitutive relations for a homogenized reinforced concrete material. Two-stage homogenization procedure is described, i.e. prior to cracking (Phase I) and after cracking (Phase II) of the concrete matrix. Hence, the localization phenomenon and the 'size effect' are properly described. The constitutive law incorporated in the main algorithm of the commercially available finite element code COSMOS/M is further discussed. The model is applied to simulate some relevant aging mechanisms. Therefore, in the proposed paper the assessment of the prestressed concrete aging of the containment structure of Gentilly-2 nuclear power plant using an advanced numerical procedure will be presented. Aging mechanisms considered possible are discussed, the present conditions are assessed and the mechanisms that are likely to impair proper future functioning of the structure are identified. The results of the numerical analysis of the reinforced concrete structure subjected to loads such as thermal and seismic loads are presented and discussed. Attention is given to the analysis of the effects of concrete swelling due to alkali-aggregate reaction. The paper also includes an evaluation of a potential damage in the context of a high velocity impact of a commercial aircraft into the containment structure. (author)

  4. Analysis of long-term behaviour of nuclear reactor containment

    Energy Technology Data Exchange (ETDEWEB)

    Hora, Z. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)]. E-mail: Zbynek.Hora@fsv.cvut.cz; Patzak, B. [Czech Technical University in Prague, Faculty of Civil Engineering, Department of Mechanics, Thakurova 7, 166 29 Prague 6 (Czech Republic)

    2007-02-15

    For assessment of safety and durability of a nuclear power plant (NPP), knowledge of the containment behaviour under various service and extreme conditions is crucial. To perform reliable analysis of such a large-scale structure, a sufficiently realistic but still feasible numerical model must be used, in which the relevant physical phenomena are reflected. Therefore, a constitutive model for concrete including effects of moisture and heat transfer, cement hydration, creep, shrinkage and optionally microcracking of concrete should be chosen. The present paper focuses on the simulation of the service life of NPP containment, aiming to determine the material and model parameters to enable reliable prediction of structural behaviour under various conditions. The purpose of the work is to provide a numerical model calibrated using existing measurements to predict the long-term behaviour reliably. Extensive in situ measurements are used to calibrate the model and to check the validity of the model hypotheses. Moreover, the material model parameters are systematically re-calibrated based on the continuous monitoring of the structure. The structural integrity test is reanalysed numerically to show the model capability of predicting behaviour of the structure under given loading and climate conditions.

  5. Modeling of thermophoretic deposition of aerosols in nuclear reactor containments

    International Nuclear Information System (INIS)

    Fernandes, A.; Loyalka, S.K.

    1996-01-01

    Aerosol released in postulated or real nuclear reactor accidents can deposit on containment surfaces via motion induced by temperature gradients in addition to the motion due to diffusion and gravity. The deposition due to temperature gradients is known as thermophoretic deposition, and it is currently modeled in codes such as CONTAIN in direct analogy with heat transfer, but there have been questions about such analogies. This paper focuses on a numerical solution of the particle continuity equation in laminar flow condition characteristics of natural convection. First, the thermophoretic deposition rate is calculated as a function of the Prandtl and Schmidt numbers, the thermophoretic coefficient K, and the temperature difference between the atmosphere and the wall. Then, the cases of diffusion alone and a boundary-layer approximation (due to Batchelor and Shen) to the full continuity equation are considered. It is noted that an analogy with heat transfer does not hold, but for the circumstances considered in this paper, the deposition rates from the diffusion solution and the boundary-layer approximation can be added to provide reasonably good agreement (maximum deviation 30%) with the full solution of the particle continuity equation. Finally, correlations useful for implementation in the reactor source term codes are provided

  6. Container materials in environments of corroded spent nuclear fuel

    Science.gov (United States)

    Huang, F. H.

    1996-07-01

    Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

  7. Building a better future: Contributions of nuclear science and technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    Food, health, energy, industry and the environment. The fields could not be more diverse. Yet they are all linked in one universe - the worldwide research and development activities of the IAEA`s Department of Research and Isotopes. These are just a few of many examples where nuclear-based technologies are helping to understand and provide constructive solutions to today`s economic and social problems. The development of these technologies through the IAEA`s Research Contract Programme and their extension through its Technical Co-operation (TC) Programme and information services encourage specialists in Member States to transform scientific results into knowledge, products and services which benefit society at large. In each of these sectors, the Department`s activities foster leadership and international cooperation in applying nuclear technology to everyday scientific, technological and socio-economic problems. Together with international meetings, they add to the collective knowledge within Member States that supports capacity building and end-user oriented development projects supported by the TC Programme

  8. Potential advantages and disadvantages of sequentially building small nuclear units instead of a large nuclear plant

    International Nuclear Information System (INIS)

    Feretic, D.; Cavlina, N.; Grgic, D.

    2008-01-01

    Renewal of nuclear power programs in countries with modest electricity consumptions and weak electrical grid interconnections has raised the question of optimal nuclear power plants sizes for such countries. The same question would be also valid for isolated or weakly connected regions within a large country. Building large size nuclear power plant could be prevented by technical or financial limits. Research programs have been initiated in the International Atomic Energy Agency and in the USA (within the framework of the Global Nuclear Energy Partnership (GNEP) program) with the aim to inspect under which circumstances small and medium reactors could be the preferred option compared to large nuclear plants. The economy of scale is a clear advantage of large plants. This paper compares, by using probabilistic methods, the net cash flow of large and medium size plants, taking as example a large nuclear plant (around 1200 MW) and four sequentially built smaller plants (300 MW). Potential advantages and disadvantageous of both options have been considered. Main advantages of the sequential construction of several identical small units could be the reduced investor risk and reduced investment costs due to the learning effect. This analysis is a part of studies for the Croatian power generating system development. (orig.)

  9. Lift-based up-ender and methods using same to manipulate a shipping container containing unirradiated nuclear fuel

    Science.gov (United States)

    Nilles, Michael J.

    2017-08-01

    A shipping container containing an unirradiated nuclear fuel assembly is lifted off the ground by operating a crane to raise a lifting tool comprising a winch. The lifting tool is connected with the shipping container by a rigging line connecting with the shipping container at a lifting point located on the shipping container between the top and bottom of the shipping container, and by winch cabling connecting with the shipping container at the top of the shipping container. The shipping container is reoriented by operating the winch to adjust the length of the winch cabling so as to rotate the shipping container about the lifting point. Shortening the winch cabling rotates the shipping container about the lifting point from a horizontal orientation to a vertical orientation, while lengthening the winch cabling rotates the shipping container about the lifting point from the vertical orientation to the horizontal orientation.

  10. Assessment of the Internal Pressure Fragility of the Hanul NPP Units 3 and 4 Containment Building Using a Nonlinear Finite Element Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyung Kui; Hahm, Dea Gi; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The sensitivity of the concrete strength is relatively higher compared to that of the steel strength. According to changes in the structure of the material, about 6-10% ultimate internal pressure differences occurred. Thirty sets of an FE model considering the material uncertainty of concrete and steel were composed for the internal pressure fragility assessment. From the internal pressure fragility assessment of the target containment building, the median capacity of liner leakage is estimated to be 116 psi. As can be seen from the Fukushima nuclear power plant accident, the containment building is the final protecting shield to prevent radiation leakage. Thus, a structural soundness evaluation for the containment pressure loads owing to a severe accident is very important. Recently, a probabilistic safety assessment has been commonly used to take into account the possible factors of uncertainty in a structural system. An assessment of the internal pressure fragility of the CANDU type containment buildings considering the correlation of structural material variables, and an assessment of the internal pressure fragility of the CANDU type containment buildings using a nonlinear finite element analysis, were also performed. However, for PWR type containment buildings, a fragility assessment has not been performed yet using a nonlinear finite element model (FEM) analysis. In this study, for the Hanul NPP units 3 and 4 containment building, the internal pressure fragility assessment was established using an FEM analysis. To do this, a three-dimensional finite element model, material property values, and a sensitive analysis were developed. A nonlinear finite element analysis of the Hanul NPP units 3 and 4 containment building was performed for a material sensitivity analysis and internal pressure fragility assessment.

  11. Assessment of the Internal Pressure Fragility of the Hanul NPP Units 3 and 4 Containment Building Using a Nonlinear Finite Element Analysis

    International Nuclear Information System (INIS)

    Park, Hyung Kui; Hahm, Dea Gi; Choi, In Kil

    2013-01-01

    The sensitivity of the concrete strength is relatively higher compared to that of the steel strength. According to changes in the structure of the material, about 6-10% ultimate internal pressure differences occurred. Thirty sets of an FE model considering the material uncertainty of concrete and steel were composed for the internal pressure fragility assessment. From the internal pressure fragility assessment of the target containment building, the median capacity of liner leakage is estimated to be 116 psi. As can be seen from the Fukushima nuclear power plant accident, the containment building is the final protecting shield to prevent radiation leakage. Thus, a structural soundness evaluation for the containment pressure loads owing to a severe accident is very important. Recently, a probabilistic safety assessment has been commonly used to take into account the possible factors of uncertainty in a structural system. An assessment of the internal pressure fragility of the CANDU type containment buildings considering the correlation of structural material variables, and an assessment of the internal pressure fragility of the CANDU type containment buildings using a nonlinear finite element analysis, were also performed. However, for PWR type containment buildings, a fragility assessment has not been performed yet using a nonlinear finite element model (FEM) analysis. In this study, for the Hanul NPP units 3 and 4 containment building, the internal pressure fragility assessment was established using an FEM analysis. To do this, a three-dimensional finite element model, material property values, and a sensitive analysis were developed. A nonlinear finite element analysis of the Hanul NPP units 3 and 4 containment building was performed for a material sensitivity analysis and internal pressure fragility assessment

  12. CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

    1993-01-01

    The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

  13. Effects of embedment including slip and separation on seismic SSI response of a nuclear reactor building

    International Nuclear Information System (INIS)

    Saxena, Navjeev; Paul, D.K.

    2012-01-01

    Highlights: ► Both the slip and separation of reactor base reduce with increase in embedment. ► The slip and separation become insignificant beyond 1/4 and 1/2 embedment respectively. ► The stresses in reactor reduce significantly upto 1/4 embedment. ► The stress reduction with embedment is more pronounced in case of tensile stresses. ► The modeling of interface is important beyond 1/8 embedment as stresses are underestimated otherwise. - Abstract: The seismic response of nuclear reactor containment building considering the effects of embedment, slip and separation at soil–structure interface requires modeling of the soil, structure and interface altogether. Slip and separation at the interface causes stress redistribution in the soil and the structure around the interface. The embedment changes the dynamic characteristics of the soil–structure system. Consideration of these aspects allows capturing the realistic response of the structure, which has been a research gap and presented here individually as well as taken together. Finite element analysis has been carried out in time domain to attempt the highly nonlinear problem. The study draws important conclusions useful for design of nuclear reactor containment building.

  14. Prediction of seismic motion from contained and excavation nuclear detonations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, R A [Environmental Research Corp., Alexandria, VA (United States)

    1970-05-15

    Capability to predict ground motions from nuclear events is developed on empirical and theoretical bases. Analyses of the experimental data provide basic predictions of peak particle motions and spectra which follow a (yield){sup m} times (distance){sup -n} relationship. The exponents on yield and distance are frequency dependent and derived from experiment and theory. Theory provides a physical understanding of the phenomena which allows extrapolation to off-NTS and atypical events. For example, yield scaling theory predicts significantly higher frequency motions and consequently larger ground accelerations for overburied events such as Gasbuggy, Rulison, Wasp and Wagon Wheel. These conclusions are observed from Gasbuggy (26 kt) which generated ground accelerations comparable to a normal buried event of 200 kt. This result is important in avoiding personal injury and assessing the probability of property damage. Conversely, theory predicts lower ground accelerations and seismic efficiencies for excavation events; these effects are observed from the Cabriolet and Schooner events and consequently predicted for the Sturtevant and Yawl events. With regard to the distance exponent, scattering theory determines a distance exponent which predicts greater attenuation effects on higher frequency motions. This trend is verified experimentally by regression analyses on a large number of data points which determine the distance exponent to range from -1.1 at low frequencies to -1.6 at high frequencies. Results indicate that cube root similarity scaling is not appropriate in the far field except possibly for peak particle displacements at the low frequency end of the spectrum. In addition to the source and transmission factors, current ground motion prediction techniques, on and off-NTS, take into account local site characteristics. Experimental evidence and theoretical models--layered media elastic theory, finite element modeling, and building response modeling

  15. Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Hennings, U.

    1987-01-01

    Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

  16. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  17. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  18. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    International Nuclear Information System (INIS)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu

    2016-01-01

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day -1 of air, 0.004%·day -1 of noble gas and 3.7×10 -5 %·day -1 of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m 3 ·hr -1 , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr -1 under the condition of 20 m·sec -1 wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor

  19. Numerical simulation of aircraft crash on nuclear containment structure

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, M.A., E-mail: iqbalfce@iitr.ernet.in [Department of Civil Engineering, Indian Institute of Technology Roorkee, Roorkee 247667 (India); Rai, S.; Sadique, M.R.; Bhargava, P. [Department of Civil Engineering, Indian Institute of Technology Roorkee, Roorkee 247667 (India)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer The deformation was more localised at the center of cylindrical portion. Black-Right-Pointing-Pointer The peak deflection at the junction of dome and cylinder was found to be 67 mm. Black-Right-Pointing-Pointer The peak deflection at midpoint of the cylindrical portion was found to be 88.9 mm. Black-Right-Pointing-Pointer The strain rate was found to be an important parameter to effect the deformation. Black-Right-Pointing-Pointer The model without strain rate and 290 s{sup -1} strain rate predicted very high deformations. - Abstract: Numerical simulations were carried with ABAQUS/Explicit finite element code in order to predict the response of BWR Mark III type nuclear containment against Boeing 707-320 aircraft crash. The load of the aircraft was applied using and force history curve. The damaged plasticity model was used to predict the behavior of concrete while the Johnson-Cook elasto-viscoplastic material model was used to incorporate the behavior of steel reinforcement. The crash was considered to occur at two different locations i.e., the midpoint of the cylindrical portion and the junction of dome and cylinder. The midpoint of the cylindrical portion experienced more deformation. The strain rate in the material model was varied and found to have a significant effect on the response of containment. The results of the present investigation were compared with those of the studies available in literature and a close agreement with the previous results was found in terms of maximum target deformation.

  20. Oxygen Containment System Options for Nuclear Thermal Propulsion Testing

    Data.gov (United States)

    National Aeronautics and Space Administration — All nuclear thermal propulsion (NTP) ground testing conducted in the 1950s and 1960s during the ROVER/(Nuclear Engine Rocket Vehicle Application (NERVA) program...

  1. Bonded or Unbonded Technologies for Nuclear Reactor Prestressed Concrete Containments

    International Nuclear Information System (INIS)

    Abrishami, Homayoun; Tcherner, Julia; Barre, Francis; Borgerhoff, Michael; Bumann, Urs; Calonius, Kim; Courtois, Alexis; Debattista, Jean-Marc; Gallitre, Etienne; Isard, Cedric; Elison, Oscar; Graves, Herman; Sircar, Madhumita; Huerta, Alejandro; White, Andrew; ); Jackson, Paul; Kjellin, Daniel; Lillhoek, Sofia; Louhivirta, Jari; Myllymaeki, Jukka; Vaelikangas, Pekka; Martin, Jose; Nakano, Makio; Puttonen, Jari; Rambach, Jean-Mathieu; Tarallo, Francois; Smith, Leslie; Stepan, Jan; Touret, Jean-Pierre; Varpasuo, Pentti

    2015-01-01

    OECD/NEA/CSNI Working Group on Integrity and Ageing of Components and Structures (WGIAGE) has the main mission to advance the current understanding of those aspects relevant to ensuring the integrity of structures, systems and components under design and beyond design loads, to provide guidance in choosing the optimal ways of dealing with challenges to the integrity of operating as well as new nuclear power plants, and to make use of an integrated approach to design, safety and plant life management. The work related to the risks of the loss of pre-stressing force in concrete structures has been in high priority during the activities of the concrete sub-group of WGIAGE. Therefore, the CAPS of WGIAGE: Study on post-tensioning methodologies in containments, was approved by CSNI in June 2009. In this study the two post-tensioning methodologies: bonded and un-bonded methods and their technological features are analysed. In the bonded technology, the tendon cannot slide in its duct due to the cement grouting which is injected after tensioning. In the un-bonded technology, the tendon can slide inside its duct, the corrosion protection is given by grease, wax or dry air. A key point concerning the assessment of durability and safety of prestressed concrete containments is the technology chosen for tendon protection: bonded with cement grout or un-bonded and protected by grease or soft products. The mechanical behaviour of the containment is directly influenced by the adherence of the tendons to the concrete, locally and under high stresses in case of severe accident. The bonded or un-bonded tendons of post-tensioned concrete containment of the Nuclear Power Plants have the major role of containment (balance of the pressure effect during design basis and beyond design accident). Many difficulties around the design, the construction and the in service inspection are related to the tendons. The main goal of the CAPS work was to clarify the consequences and necessary

  2. Engineering and innovative erection concept for the containment liner for a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Goehring, Rainer [Babcock Noell GmbH, Wuerzburg (Germany). Abt. Nukleare Projekte; Anders, Nils; Nagelstutz, Franz [Babcock Noell GmbH, Wuerzburg (Germany). Abt. Berechnung

    2010-04-15

    The determination of the optimal design of the containment liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. Several load cases for normal operation and accidental conditions as well as severe accidents have been analysed. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. The construction costs of a Nuclear Power Plant are also impacted by the erection time of components. Therefore, it is necessary to optimize the erection times and consequently essentially to shorten. One possibility has been demonstrated by Babcock Noell GmbH with the erection concept of a containment liner. The liner is preassembled on the pre-assembly place in rings of up to 12 meter height, with the nominal diameter of 47 meter and a weight of approximately up to 200 tonnes. These rings, as well as the containment cup and the dome, are lifted with a heavy load crane into the reactor building. With this concept, the construction activities on the inner containment wall are only five times disrupted by the welding and coating of the circumferential weld (approx. 25 calendar days). In comparison with the common known erection concept of welding of liner shells in situ, at least 20 weeks are saved on the schedule. An integrated concept from planning, manufacturing and erection of this large component has been implemented. It could be demonstrated that within the given time frame, with the required quality and within the required tolerances the containment liner for the Nuclear Power Plant can be delivered to the Client. (orig.)

  3. Engineering and innovative erection concept for the containment liner for a nuclear power plant

    International Nuclear Information System (INIS)

    Goehring, Rainer; Anders, Nils; Nagelstutz, Franz

    2010-01-01

    The determination of the optimal design of the containment liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. Several load cases for normal operation and accidental conditions as well as severe accidents have been analysed. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. The construction costs of a Nuclear Power Plant are also impacted by the erection time of components. Therefore, it is necessary to optimize the erection times and consequently essentially to shorten. One possibility has been demonstrated by Babcock Noell GmbH with the erection concept of a containment liner. The liner is preassembled on the pre-assembly place in rings of up to 12 meter height, with the nominal diameter of 47 meter and a weight of approximately up to 200 tonnes. These rings, as well as the containment cup and the dome, are lifted with a heavy load crane into the reactor building. With this concept, the construction activities on the inner containment wall are only five times disrupted by the welding and coating of the circumferential weld (approx. 25 calendar days). In comparison with the common known erection concept of welding of liner shells in situ, at least 20 weeks are saved on the schedule. An integrated concept from planning, manufacturing and erection of this large component has been implemented. It could be demonstrated that within the given time frame, with the required quality and within the required tolerances the containment liner for the Nuclear Power Plant can be delivered to the Client. (orig.)

  4. Radionuclide transport through perforations in nuclear waste containers

    International Nuclear Information System (INIS)

    Aidun, C.K.; Bloom, S.G.; Raines, G.E.

    1987-11-01

    Previous analytical models for the steady-state radionuclide release rate through perforations in nuclear waste containers into the surrounding medium are based on a zero wall thickness assumption. In this paper we investigate the effect of the wall thickness on the mass transfer rate through isolated cylindrical holes. We solve the steady-state diffusion equation for the concentration field and derive a model based on the analytical solution. By direct comparison, we show that the zero wall thickness model overpredicts the mass transfer rate by about 1300 percent for a circular hole with 1-cm radius and a wall thickness of 10 cm. As expected, the zero-thickness model becomes even less accurate as the hole radius decreases; it predicts a greater release rate from a large number of small holes than the mass transfer rate from an uncontained waste form cylinder. In contrast, the results predicted by our model remain bounded for isolated holes and never exceed the mass transfer from an uncontained waste form. 6 refs., 9 figs., 3 tabs

  5. Characterization of nuclear reactor containment penetrations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Shackelford, M.H.; Bump, T.R.; Seidensticker, R.W.

    1985-02-01

    This report concludes a preliminary report prepared by ANL for Sandia, published as NUREG/CR-3855, in June 1984. The preliminary report, NUREG/CR-3855, presented the results of a survey of nuclear reactor containment penetrations, covering the number of plants surveyed at that time (22 total). Since that time, an additional 26 plants have been included in the survey. This final report serves two purposes: (1) to add the summary data sheets and penetration details for the additional plants now included in the survey; and (2) to confirm, revise, or add to analyses and discussions presented in the first report which, of course, were based solely on the earlier sample of 22 plants. This final report follows the outline and format of the preliminary survey report. In general, changes and additions to the preliminary report are implied, rather than stated as such to avoid repeated reference to that report. If no changes have been made in a section the title of the section of the previous report is simply repeated followed by ''No Changes''. Some repetition is used for continuity and clarity.

  6. An insight on the nuclear competence building pillars

    Energy Technology Data Exchange (ETDEWEB)

    Wieland, P. [Comissao Nacional de Energia Nuclear, Rua General eertano, 90 - 22.290-901 - Rio de Janeiro (Brazil)]. e-mail: pwieland@cnen.gov.br

    2007-07-01

    knowledge if one is willing to collaborate in a highly motivated way, and the other one is passive or mentally absent. This paper takes an insight on competence building from a sensible point of view, sharing practical experiences from the nuclear area. (Author)

  7. An insight on the nuclear competence building pillars

    International Nuclear Information System (INIS)

    Wieland, P.

    2007-01-01

    knowledge if one is willing to collaborate in a highly motivated way, and the other one is passive or mentally absent. This paper takes an insight on competence building from a sensible point of view, sharing practical experiences from the nuclear area. (Author)

  8. The role of inertial containment fusion in replacing nuclear tests

    Energy Technology Data Exchange (ETDEWEB)

    Schaper, Annette [Hessische Stiftung Friedens- und Konfliktforschung, Frankfurt am Main (Germany)

    2008-07-01

    Nuclear weapon physicists need to understand the process of a nuclear explosion, and their major experimental tools had been nuclear tests. Since a couple of years, the established nuclear weapon states observe a testing moratorium. Nevertheless, they still want to keep their nuclear arsenals, and consequently to ensure the reliability, safety, and security of their nuclear warheads. For this purpose, they use experimental tools that replace nuclear tests, among them ICF. ICF plays a central role in the so-called ''stockpile stewardship program'' that the U.S. has implemented when it participated in the negotiations on a Comprehensive Test Ban Treaty. Several questions arise and are discussed in the presentation: Does ICF allow to simulate the extreme conditions of a nuclear explosion? Which are the functions of nuclear testing that ICF can replace and which are beyond its capabilities? Would ICF be a useful tool for the design of new nuclear warheads? Why are so huge sums spent on ICF in a military context although the usefulness for nuclear weapons seems rather limited?.

  9. Proceedings of the 25th Seismic Research Review -- Nuclear Explosion Monitoring: Building the Knowledge Base

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, Francesca C. [Editor; Mendius, E. Louise [Editor

    2003-09-23

    These proceedings contain papers prepared for the 25th Seismic Research Review -- Nuclear Explosion Monitoring: Building the Knowledge Base, held 23-25 September, 2003 in Tucson, Arizona. These papers represent the combined research related to ground-based nuclear explosion monitoring funded by the National Nuclear Security Administration (NNSA), Defense Threat Reduction Agency (DTRA), Air Force Research Laboratory (AFRL), US Army Space and Missile Defense Command, and other invited sponsors. The scientific objectives of the research are to improve the United States capability to detect, locate, and identify nuclear explosions. The purpose of the meeting is to provide the sponsoring agencies, as well as potential users, an opportunity to review research accomplished during the preceding year and to discuss areas of investigation for the coming year. For the researchers, it provides a forum for the exchange of scientific information toward achieving program goals, and an opportunity to discuss results and future plans. Paper topics include: seismic regionalization and calibration; detection and location of sources; wave propagation from source to receiver; the nature of seismic sources, including mining practices; hydroacoustic, infrasound, and radionuclide methods; on-site inspection; and data processing.

  10. Proceedings of the 25th Seismic Research Review -- Nuclear Explosion Monitoring: Building the Knowledge Base

    International Nuclear Information System (INIS)

    Chavez, Francesca C.; Mendius, E. Louise

    2003-01-01

    These proceedings contain papers prepared for the 25th Seismic Research Review -- Nuclear Explosion Monitoring: Building the Knowledge Base, held 23-25 September, 2003 in Tucson, Arizona. These papers represent the combined research related to ground-based nuclear explosion monitoring funded by the National Nuclear Security Administration (NNSA), Defense Threat Reduction Agency (DTRA), Air Force Research Laboratory (AFRL), US Army Space and Missile Defense Command, and other invited sponsors. The scientific objectives of the research are to improve the United States capability to detect, locate, and identify nuclear explosions. The purpose of the meeting is to provide the sponsoring agencies, as well as potential users, an opportunity to review research accomplished during the preceding year and to discuss areas of investigation for the coming year. For the researchers, it provides a forum for the exchange of scientific information toward achieving program goals, and an opportunity to discuss results and future plans. Paper topics include: seismic regionalization and calibration; detection and location of sources; wave propagation from source to receiver; the nature of seismic sources, including mining practices; hydroacoustic, infrasound, and radionuclide methods; on-site inspection; and data processing.

  11. Extreme accident mitigation - analysis of a low pressure secondary containment building

    International Nuclear Information System (INIS)

    Vaughan, G.J.; Dunbar, I.H.

    1987-01-01

    Although whole core accidents are sufficiently unlikely as to be beyond the design basis, the Secondary Containment Building [SCB] is expected to have some effect in mitigating the consequences of such accidents. From a design point of view there are many advantages in having a low pressure SCB fitted with a filtered vent, so studies have been undertaken of the response of such a building to the large sodium fires that might follow a severe accident. The behaviour of the sodium oxide aerosols has been studied using the code AEROSIM. The efficiency of an aerosol scrubber has been investigated experimentally. A simple code, SECCONTAIN, has been developed to model the effects of sodium fires in buildings, and has been applied to a specific design of a low pressure SCB. (author)

  12. The nuclear waste containment and some aspects of the deep disposal concept

    International Nuclear Information System (INIS)

    Felix, B.; Thorner, P.; Raimbault, P.; Beaulieu, F.

    1995-01-01

    The French agency for the management of nuclear waste, ANDRA, is in charge of investigating the feasibility of deep disposal of high level waste in at least two types of geologic formation, leading to the validation of disposal concepts with and without retrievability. Plans to build two underground laboratories are afoot. Meanwhile, parametric modelling studies have been performed, with interesting results, some of which are shown here in graphic form. It is proved that if the overpack surrounding waste containers can be made to last for a thousand years, the dose resulting from Sr-90 and Cs-137 is nil. Conversely, the dose from actinides such as americium and Th-229 is largely unaffected by the package, being determined by their own low solubilities and underground water flow. Temperature rise in a granite host formation was modelled as a function of the distance between disposal boreholes. Finite element two dimensional calculations of water flow through backfill were also performed. 1 ref., 8 figs

  13. Ultimate capacity and influenced factors analysis of nuclear RC containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Song Chenning; Hou Gangling; Zhou Guoliang

    2014-01-01

    Ultimate compressive bearing capacity, influenced factors and its rules of nuclear RC containment are key problems of safety assessment, accident treatment and structure design, etc. Ultimate compressive bearing capacity of nuclear RC containment is shown by concrete damaged plasticity model and steel double liner model of ABAQUS. The study shows that the concrete of nuclear RC containment cylinder wall becomes plastic when the internal pressure is up to 0.87 MPa, the maximum tensile strain of steel liner exceeds 3000 × 10 6 and nuclear RC containment reaches ultimate status when the internal pressure is up to 1.02 MPa. The result shows that nuclear RC containment is in elastic condition under the design internal pressure and the bearing capacity meets requirement. Prestress and steel liner play key parts in the ultimate internal pressure and failure mode of nuclear RC containment. The study results have value for the analysis of ultimate compressive bearing capacity, structure design and safety assessment. (authors)

  14. Experimental study of the leakage rate through cracked reinforced concrete wall elements for defining the functional failure criteria of containment buildings

    International Nuclear Information System (INIS)

    Choun, Young Sun; Cho, Nam So

    2004-01-01

    Containment buildings in nuclear power plants should maintain their structural safety as well as their functional integrity during an operation period. To maintain the functional integrity, the wall and dome of the containment buildings have to maintain their air tightness under extreme loading conditions such as earthquakes, missile impact, and severe accidents. For evaluating the functional failure of containments, it is important to predict the leak amount through cracked concrete walls. The leakage through concrete cracks has been studied since 1972. Buss examined the flow rate of air through a pre-existing crack in a slab under air pressure. Rizkalla el al. initiated an experimental study for the leakage of prestressed concrete building segments under uniaxial and biaxial loadings to simulate the loading condition of containment buildings under an internal pressure. Recently, Salmon el al. initiated an experimental program for determining the leak rates in typical reinforced concrete shear walls subjected to beyond design basis earthquakes. This study investigates the cracking behavior of reinforced concrete containment wall elements under a uniaxial tension and addresses the outline of the leakage test for unlined containment wall elements

  15. Toward a common nuclear safety culture: from knowledge creation to competence building in Euratom programmes

    International Nuclear Information System (INIS)

    Van Goethem, G.

    2010-01-01

    Content of the presentation: Introduction: towards a common nuclear safety culture 2. EU Stakeholders in nuclear fission and « Nuclear Safety Directive » June 2009 3. EURATOM policy for education (from knowledge creation …) 4. EURATOM policy for training (… to competence building) 5. Examples of EFTS running under FP-7 EURATOM 6. Conclusion: EC “seed money” for effort shared with MS

  16. Discussion on building safety culture inside a nuclear safety regulatory body

    International Nuclear Information System (INIS)

    Fan Yumao

    2013-01-01

    A strong internal safety culture plays a key role in improving the performance of a nuclear regulatory body. This paper discusses the definition of internal safety culture of nuclear regulatory bodies, and explains the functions that the safety culture to facilitate the nuclear safety regulation and finally puts forward some thoughts about building internal safety culture inside regulatory bodies. (author)

  17. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  18. Stress corrosion cracking of candidate materials for nuclear waste containers

    International Nuclear Information System (INIS)

    Maiya, P.S.; Shack, W.J.; Kassner, T.F.

    1989-09-01

    Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

  19. Dose and radon measurements inside houses containing ash as building material

    International Nuclear Information System (INIS)

    Bodnar, R.; Lendvai, Z.; Somlai, J.; Nemeth, C.

    1996-01-01

    Radon concentration and external dose have been measured in dwellings that contain by-products of coal burning for building materials. The concentrations of 40 K, 232 Th, 238 U and 226 Ra have been determined in the materials. The date are analyzed according to indices frequently used for decision of utilizing the by-products. The observed daily fluctuation of the radon concentration in dwellings might exceed a factor of 5. (author)

  20. Evaluation of Shear Resisting Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    International Nuclear Information System (INIS)

    Choun, Youngsun; Park, Junhee

    2014-01-01

    Conventional reinforced concrete (RC) members generally show a rapid deterioration in shear resisting mechanisms under a reversed cyclic load. However, the use of high-performance fiber-reinforced cement composites provides excellent damage tolerance under large displacement reversals compared with regular concrete. Previous experimental studies have indicated that the use of fibers in conventional RC can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. This study evaluates the shear resisting capacity for a PCCB constructed using steel fiber reinforced concrete (SFRC) or polyamide fiber reinforced concrete (PFRC). The effects of steel and polyamide fibers on the shear performance of a PCCB were investigated. It was revealed that steel fibers are more effective to enhance the shear resisting capacity of a PCCB than polyamide fibers. The ductility and energy dissipation increase significantly in fiber reinforced PCCBs

  1. Evaluation of Shear Resisting Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Park, Junhee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Conventional reinforced concrete (RC) members generally show a rapid deterioration in shear resisting mechanisms under a reversed cyclic load. However, the use of high-performance fiber-reinforced cement composites provides excellent damage tolerance under large displacement reversals compared with regular concrete. Previous experimental studies have indicated that the use of fibers in conventional RC can enhance the structural and functional performance of prestressed concrete containment buildings (PCCBs) in nuclear power plants. This study evaluates the shear resisting capacity for a PCCB constructed using steel fiber reinforced concrete (SFRC) or polyamide fiber reinforced concrete (PFRC). The effects of steel and polyamide fibers on the shear performance of a PCCB were investigated. It was revealed that steel fibers are more effective to enhance the shear resisting capacity of a PCCB than polyamide fibers. The ductility and energy dissipation increase significantly in fiber reinforced PCCBs.

  2. Building trust with the schoolchildren in the nuclear training centre

    International Nuclear Information System (INIS)

    Stritar, Andrej; Istenie, Radko

    1995-01-01

    Although Krsko Nuclear Power Plant has been in operation for more than ten years, comparatively little has been done in the field of systematic education and public information. Deficiencies in this field are causing serious misunderstandings about the role of nuclear power and is having a negative impact on its public acceptance. At The Nuclear Training Centre Milan Copic (NTC) in Ljubljana we have prepared a presentation on nuclear energy encompassing a short description of a PWR nuclear power plant, importance of nuclear power in the world and in Slovenia, basic ideas of nuclear safety and radioactive waste disposal

  3. The analysis of the containment building for global effects of an aircraft crash

    International Nuclear Information System (INIS)

    Varpasuo, P.; Kenttaelae, J.

    1981-01-01

    The aim of the analysis was to establish the displacement and stress states for the whole building in four time points during the transient as well as determine the acceleration time histories, and acceleration response spectra for five points inside the containment building. The five points inside the containment were located between the foundation slab and the upper edge of the containment building. The total amount of nodal points in the model was 800 and the total amount of degrees of freedom was 4300; the amount of quadrilateral shell elements was 900. STARDYNE AND SAPOV programs and direct integration were used for analysis. As for the results of the analysis the following statements can be made: The results calculated by SAPIV and STARDYNE programs are essentially identical. Only the more refined LCCT-11 shell element used in STARDYNE instead of LCCT-9 element used in SAPIV causes the STARDYNE model to be more flexible and this shows in the response histories as a certain phase lag. STARDYNE response leaves gradually behind SAPIV response. The unexpected thing in acceleration responses was that the response in horizontal direction perpendicular to the load application direction was twice as large as the acceleration response in load application direction. (orig./HP)

  4. LEAF: a computer program to calculate fission product release from a reactor containment building for arbitrary radioactive decay chains

    International Nuclear Information System (INIS)

    Lee, C.E.; Apperson, C.E. Jr.; Foley, J.E.

    1976-10-01

    The report describes an analytic containment building model that is used for calculating the leakage into the environment of each isotope of an arbitrary radioactive decay chain. The model accounts for the source, the buildup, the decay, the cleanup, and the leakage of isotopes that are gas-borne inside the containment building

  5. New CSA guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances

    International Nuclear Information System (INIS)

    Rhodes, M.; Kwong, A.

    2011-01-01

    The Canadian Standards Association (CSA) guideline N292.5, Guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances, was recently developed to address a need for guidance on approaches for clearance of materials from facilities licensed by the Canadian Nuclear Safety Commission (CNSC) consistent with Canadian and international recommendations. This guideline is also applicable to determining if an activity associated with materials that contain nuclear substances is exempt from requiring a CNSC licence. The guideline summarizes the regulatory requirements associated with the exemption and clearance of materials and provides a graded approach to designing a survey based on the risk of residual contamination being present. (author)

  6. Emergency cooling process and device for nuclear reactor containment

    International Nuclear Information System (INIS)

    Costes, D.

    1985-01-01

    The emergency cooling system of a PWR containment, according to the principal patent, comprises a turbine fed by the humid air of the containment, a condenser in which the air flowing out of the turbine is dryed and cooled by an external coolant and a compressor actuated by the turbine and returning the dryed air in the containment [fr

  7. Nuclear Science Capacity Building in Kenya: Challenges and Opportunities

    International Nuclear Information System (INIS)

    Mangala, J. M.

    2017-01-01

    Kenya's significant involvement in Nuclear Science and Technology can be traced back to 1965 when the country became a member state of the International Atomic Energy Agency (IAEA). In 1978, Kenya formulated a project for the establishment of the ''Nuclear Science Laboratory'' at the University of Nairobi that soon after, received assistance from International Atomic Energy Agency. The laboratory was expected to be a base for the promotion of nuclear science technologies in the country was started in 1979 and has since developed into a fully-fledged institute of the University of Nairobi. In general, six main areas of nuclear science applications have continued to receive IAEA assistance; during the past ten years ; agriculture and soil management (30%), livestock production , introduction to nuclear power production (21%)- radiation oncology in cancer management and nuclear medicine (16%). Smaller shares went to nuclear safety (9%), nuclear engineering and technology (8%), industry and water resource management (7%) and nuclear physics and chemistry (5%). At present, the Agency is supporting several technical co-operation projects, four of which are in agriculture and two in nuclear physics and chemistry with additional assistance in the areas of manpower development, nuclear medicine, non-destructive testing techniques and radioactive waste management. Thus, through Government initiatives, and with the assistance of IAEA, quite a number of specialist national laboratories for nuclear science application have emerged

  8. North Korea’s 2009 Nuclear Test: Containment, Monitoring, Implications

    Science.gov (United States)

    2010-04-02

    inspections as prima facie evidence of a violation. One generally-accepted means of evading detection of nuclear tests, especially low-yield tests...In an attempt to extend these bans to cover all nuclear tests, negotiations on the CTBT were completed in 1996. The treaty’s basic obligation is to...Verification refers to determining whether a nation is in compliance with its treaty obligations , which in this case means determining whether a suspicious

  9. Forced vibration tests on the reactor building of a nuclear power station, 1

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Tsunoda, Tomohiko; Wakamatsu, Kunio; Kaneko, Masataka; Nakamura, Mitsuru; Kunoh, Toshio; Murahashi, Hisahiro

    1988-01-01

    Tsuruga Unit No.2 Nuclear Power Station of the Japan Atomic Power Company is the first PWR-type 4-loop plant constructed in Japan with a prestressed concrete containment vessel (PCCV). This report describes forced vibration tests carried out on the reactor building of this plant. The following were obtained as results: (1) The results of the forced vibration tests corresponded well on the whole with design values. (2) The vibration characteristics of the PCCV observed in the tests after prestressing are no different from the ones before prestressing. This shows that the vibration properties of the PCCV are practically independent of prestressing loads. (3) A seismic response analysis of the design basis earthquake was made on the design model reflecting the test results. The seismic safety of the plant was confirmed by this analysis. (author)

  10. Terrain and building effects on the transport of radioactive material at a nuclear site

    International Nuclear Information System (INIS)

    Jeong, Hyojoon; Park, Misun; Jeong, Haesun; Hwang, Wontae; Kim, Eunhan; Han, Moonhee

    2014-01-01

    Highlights: • This study is to quantify the building and terrain effects on the atmospheric dispersion. • Statistical methods with AERMOD-PRIME and CFD were used. • To assess the risk in nuclear power plants, terrain and building effects have to be considered. - Abstract: This study identified the terrain and building effects on the atmospheric dispersion of radioactive materials at the Wolsong Nuclear Site. To analyze the atmospheric dispersion of radioactive materials, the AERMOD-PRIME model, CFD model and meteorological data from 2010 were used. The terrain and building effects on the atmospheric dispersion of radioactive materials within a 1 km radius of the site were statistically significant. The maximum concentration of the radioactive material increased by 7 times compared to the concentration when the terrain and building effects were not considered. It was found that the terrain and building influenced the decrease in the concentration of radioactive material in a concentric circle with a 914 m radius from the center of the site. The concentration of radioactive material in a concentric circle with a 350 m radius was two-times higher than the concentration estimated at the backside of the building, which is the downwind side, without any consideration of the terrain and building effects. In consideration of the Korean situation, in which multiple nuclear reactors are built on the same nuclear site, it is necessary to evaluate the risk that may affect workers and nearby residents by reflecting the terrain and building effects

  11. Research programme related to the influence of wind on contamination containment in nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Soares, S.; Gelain, T.; Laborde, J.C.; Ricciardi, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN/DSU/SERAC/LEMAC), 91 - Gif-sur-Yvette (France)

    2006-07-01

    In nuclear industry, the response of a ventilation network to accidental disturbances, either mechanical (fan failure, damper blockage,..) or thermal (fire..) is difficult to evaluate when the network becomes complex. In order to determine and analyze the consequences of these disturbances on the radioactive materials containment, a computer code called SIMEVENT has been developed. However, among the external parameters likely to affect a ventilation network, the wind effect is actually basically modeled, due to a lack of qualified data concerning the wind impact on complex building geometries and the interaction between wind and chimney exhaust. In view of the networks complexity and the facilities diversity, a research program including experimental and model studies has been launched to assess the wind influence on contamination containment. 1. step: improvement of data (2005-2006): The diversity of facilities geometries needs the use of a qualified multi-D code for pressure coefficients Cpi assessment, characterizing the wind effect on building walls and the interaction between wind and chimney exhaust. Different chimney terminals have then been placed in a wind tunnel (the parameters are the incline angle a, the wind velocity U and the air flow in the duct W); for each angle, the evolution of the pressure coefficient versus wind velocity is determined and is characteristic of a chimney terminal geometry. Furthermore, two types of scale-model have been chosen for representing either nuclear power plants (NPP) or plants and laboratories buildings. The different values of wind pressure coefficients have been measured on both scale-models placed in a wind-tunnel. The experimental data obtained are compared with CFD simulations (CFX code), in order to qualify such code for the assessment of pressure coefficient on complex geometries. The results are quite encouraging. 2. step - wind tunnel tests on a ventilated scale model (2007-2009): Wind tunnel tests will be

  12. Research programme related to the influence of wind on contamination containment in nuclear installations

    International Nuclear Information System (INIS)

    Soares, S.; Gelain, T.; Laborde, J.C.; Ricciardi, L.

    2006-01-01

    In nuclear industry, the response of a ventilation network to accidental disturbances, either mechanical (fan failure, damper blockage,..) or thermal (fire..) is difficult to evaluate when the network becomes complex. In order to determine and analyze the consequences of these disturbances on the radioactive materials containment, a computer code called SIMEVENT has been developed. However, among the external parameters likely to affect a ventilation network, the wind effect is actually basically modeled, due to a lack of qualified data concerning the wind impact on complex building geometries and the interaction between wind and chimney exhaust. In view of the networks complexity and the facilities diversity, a research program including experimental and model studies has been launched to assess the wind influence on contamination containment. 1. step: improvement of data (2005-2006): The diversity of facilities geometries needs the use of a qualified multi-D code for pressure coefficients Cpi assessment, characterizing the wind effect on building walls and the interaction between wind and chimney exhaust. Different chimney terminals have then been placed in a wind tunnel (the parameters are the incline angle a, the wind velocity U and the air flow in the duct W); for each angle, the evolution of the pressure coefficient versus wind velocity is determined and is characteristic of a chimney terminal geometry. Furthermore, two types of scale-model have been chosen for representing either nuclear power plants (NPP) or plants and laboratories buildings. The different values of wind pressure coefficients have been measured on both scale-models placed in a wind-tunnel. The experimental data obtained are compared with CFD simulations (CFX code), in order to qualify such code for the assessment of pressure coefficient on complex geometries. The results are quite encouraging. 2. step - wind tunnel tests on a ventilated scale model (2007-2009): Wind tunnel tests will be

  13. Theoretical investigations to the mass- and energy transport in light water reactor containment buildings

    International Nuclear Information System (INIS)

    Hocke, K.D.; Bisanz, R.; Ploeger, G.

    1982-12-01

    In this report loss-of-coolant accidents in water-cooled nuclear power reactors will be described with regard to their effects on the reactor containment system. The general presentation of containment systems and the accident sequences is completed by performances of the physical and mathematical treatment of the flow dynamics. Following a survey of existing computer codes for containment analysis problems, a detailed model description of the computer codes ZOCO 6 and BEACON-MOD 2A is given. A variation of important model parameters within the ZOCO 6 program explains the field of application and leads to a valuation of the model assumptions. The obtained results are presented graphically including latest publications. (orig.) [de

  14. Model of the containment building of Almaraz NPP and the system of recombiners PARs, with the GOTHIC code, for the study of the diffusion of combustible gases

    International Nuclear Information System (INIS)

    Garcia Gonzalez, M.; Huelamo, E.; Mazrtinez, M.; Perez, J. R.

    2014-01-01

    This paper presents the analysis of distribution of gases within the containment building carried out a simulation model with the code Thermo hydraulic GOTHIC, which has been evaluated based on passive autocatalytic recombiners gas control system. The model considers scenarios of severe accident with specific conditions that produce the most hydrogen generation rates. Intended to verify the effectiveness of the control system of gas expected to be installed in the Almaraz Nuclear power plant so that the number and location of recombiners equipment meets its function of preventing the formation of explosive atmospheres which impairs the integrity of the containment, reducing and limiting the concentration of combustible gases during the postulated accident. (Author)

  15. The elementary study on the building of the virtual nuclear monitoring system

    International Nuclear Information System (INIS)

    Tang Wanlin; Dai Zhuangrong; Zhang Yu

    2004-01-01

    In this paper we discuss the feasibility to build the virtual nuclear monitoring system using the modern detecting technology, the modern information managing technology and the computer virtual reality technology. We also present the model of this system. (authors)

  16. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  17. Aging of steel containments and liners in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Norris, W.E.

    1998-02-01

    Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation

  18. Preliminary research on time degradation of mechanical characteristics of concretes used in nuclear power plant buildings

    International Nuclear Information System (INIS)

    Ciornei, R.

    1991-01-01

    To provide severe safety rules governing the operation of nuclear power plants, reinforced and concrete elements and structures should preserve the quality and time-constant parameters throughout the life-time of the buildings. Some important design parameters are concrete strength and elasticity modulus. Preliminary research on concrete specimens made in laboratory whose strength and static and dynamic elasticity modulus have been determined after an ageing test, has aimed at nuclear power design and building. (author)

  19. Main building of the Paks nuclear power plant, Hungary

    International Nuclear Information System (INIS)

    Fejes, A.

    1983-01-01

    The general layout of the main service building of the power plant, the applied building materials as well as the prefabricated structures are described. The conditions of planning and construction are discussed. Novel construction methods under the given conditions were utilized. (author)

  20. Processing of nuclear power plant waste streams containing boric acid

    International Nuclear Information System (INIS)

    1996-10-01

    Boric acid is used in PWR type reactor's primary coolant circuit to control the neutron flux. However, boric acid complicates the control of water chemistry of primary coolant and the liquid radioactive waste produced from NPP. The purpose of this report is to provide member states with up-to-date information and guidelines for the treatment and conditioning of boric acid containing wastes. It contains chapters on: (a) characteristics of waste streams; (b) options for management of boric acid containing waste; (c) treatment/decontamination of boric acid containing waste; (d) concentration and immobilization of boric acid containing waste; (e) recovery and re-use of boric acid; (f) selected industrial processes in various countries; and (g) the influence of economic factors on process selection. 72 refs, 23 figs, 5 tabs

  1. Gas flow in and out of a nuclear waste container

    International Nuclear Information System (INIS)

    Zwahlen, E.D.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-05-01

    We analyze the flow of gases out of and into a high-level-waste container in the unsaturated tuff of Yucca Mountain. Containers are expected to fail eventually by localized cracks and penetrations. Even though the penetrations may be small, argon gas initially in the hot container can leak out. As the waste package cools, the pressure inside the container can become less than atmospheric, and air can leak in. 14 C released from the hot fuel-cladding surface can leak out of penetrations, and air inleakage can mobilize additional 14 C and other volatile radioactive species as it oxidizes the fuel cladding and the spent fuel. In an earlier paper we studied the gas flow through container penetrations occurring at the time of emplacement. Here we analyze the flow of gas for various penetration sizes occurring at 300 years. 3 refs., 2 figs

  2. The ''nuclear car wash'': a scanner to detect illicit special nuclear material in cargo containers

    International Nuclear Information System (INIS)

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Dougan, A. D.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Pohl, B. A.; Prussin, S. G.; Walling, R. S.; Weirup, D. L.

    2004-01-01

    There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for detection of fissionable material hidden in maritime cargo containers. The system makes use of a low intensity neutron beam for producing fission; and the detection of the abundant high-energy γ rays emitted in the β-decay of short-lived fission products and β-delayed neutrons. The abundance of the delayed γ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission and they are emitted on about the same time scale as the delayed neutrons, i.e., ∼1 min. The energy and temporal distributions of the delayed γ rays provide a unique signature of fission. Because of their high energy, these delayed γ rays penetrate loW--Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed γ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The γ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5 MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source

  3. Can we ever build lots of new nuclear plants?

    International Nuclear Information System (INIS)

    Kidd, St.

    2004-01-01

    The order to construct the fifth high-power Finnish reactor is both a significant even for the West-European nuclear power sector and a promising event for the development of the world nuclear power sector. One considers the peculiar features of the economic situation in Finland as compared with other countries. One underlines that to ensure the competitive power of new nuclear power units in the liberalized energy markets one shall ensure essential reduction of the capital costs [ru

  4. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  5. Multi-barrier containment for nuclear power plants

    International Nuclear Information System (INIS)

    Koehler, A.C.; Leppke, D.M.; Corcoran, R.W.

    1968-01-01

    The multi-barrier containment concept and the basic structural design citeria for its adaptation to PWR systems are described. Some of the significant factors that led to the acceptance of the concept are also discussed

  6. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  7. Introduction to the controlled nuclear fusion (magnetic containment systems)

    International Nuclear Information System (INIS)

    Cabrera, J.A.; Guasp, J.; Martin, R.

    1975-01-01

    The magnetic containment systems, their more important features, and their potentiality to became thermonuclear reactors is described. The work is based upon the first part of a set of lectures dedicated to Plasma and Fusion Physics. (author)

  8. Locks at reactor safety containments of nuclear power plants

    International Nuclear Information System (INIS)

    1978-12-01

    The rule is to be applied for material access locks. A material access lock as defined by this rule is a pressure resistant and technically gas-tight hollow body that is connected with the containment and has got two doors, the inner door of which connects the lock chamber with the interior of the containment and the outer door connects the lock chamber with the outside. It serves exclusively for transferring material or objects. (orig./HP) [de

  9. Airlocks in the containment of nuclear power plants

    International Nuclear Information System (INIS)

    1978-01-01

    An equipment hatch according to this regulator is a pressure-resisting and technically gas-tight hollow body which is connected with the containment and has got two doors, the inner door connecting the inner space of the hatch with the interior of the containment and the outer door connecting the inner space of the hatch with the exterior. It is exclusively destinied for the transport of material or objects and not of persons. (orig./HP) [de

  10. Education and Training, and Knowledge Networks for Capacity-Building in Nuclear Security

    International Nuclear Information System (INIS)

    Mrabit, Khammar

    2014-01-01

    Conclusions: • Capacity Building (CB) is critical for States to establish and maintain effective and sustainable nuclear security regime. • IAEA is a worldwide platform promoting international cooperation for CB in nuclear security involving more than 160 countries and over 20 Organizations and Initiatives. • IAEA Division of Nuclear Security is ready to continue supporting States in developing their CB through: – Comprehensive Training Programme: more than 80 training events annually – International Nuclear Security Training and Support Centre Network (NSSC) – Comprehensive Education Programme – International Nuclear Security Network (INSEN)

  11. Integrated leak rate testing of the fast flux test facility reactor containment building

    International Nuclear Information System (INIS)

    James, E.B.; Farabee, O.A.; Bliss, R.J.

    1978-01-01

    The initial Integrated Leak Rate Test (ILRT) of the Fast Flux Test Facility containment building was performed from May 27 to June 2, 1978. The test was conducted in air with systems vented and with the containment recirculating coolers in operation. 10 psig and 5 psig tests were run using the absolute pressure test method. The measured leakage rates were .033% Vol/24 hr. and -.0015% Vol/24 hrs. respectively. Subsequent verification tests at both 10 psig and 5 psig proved that the test equipment was operating properly and it was sensitive enough to detect leaks at low pressures. This ILRT was performed at a lower pressure than any previous ILRT on a reactor containment structure in the United States. While the initial design requirements for ice condenser containments called for a part pressure test at 6 psig, the tests were waived due to the apparent statistical problems of data analysis and the repeatability of the data itself at such low pressure. In contrast to this belief, both the 5 and 10 psig ILRT's were performed in a successful manner at FFTF

  12. Steam explosions-induced containment failure studies for Swiss nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

    1998-01-01

    The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

  13. Corrosion issues in high-level nuclear waste containers

    Science.gov (United States)

    Asl, Samin Sharifi

    In this dissertation different aspects of corrosion and electrochemistry of copper, candidate canister material in Scandinavian high-level nuclear waste disposal program, including the thermodynamics and kinetics of the reactions that are predicted to occur in the practical system have been studied. A comprehensive thermodynamic study of copper in contact with granitic groundwater of the type and composition that is expected in the Forsmark repository in Sweden has been performed. Our primary objective was to ascertain whether copper would exist in the thermodynamically immune state in the repository, in which case corrosion could not occur and the issue of corrosion in the assessment of the storage technology would be moot. In spite of the fact that metallic copper has been found to exist for geological times in granitic geological formations, copper is well-known to be activated from the immune state to corrode by specific species that may exist in the environment. The principal activator of copper is known to be sulfur in its various forms, including sulfide (H2S, HS-, S2-), polysulfide (H2Sx, HSx -, Sx 2-), poly sulfur thiosulfate ( SxO3 2-), and polythionates (SxO6 2-). A comprehensive study of this aspect of copper chemistry has never been reported, and yet an understanding of this issue is vital for assessing whether copper is a suitable material for fabricating canisters for the disposal of HLNW. Our study identifies and explores those species that activate copper; these species include sulfur-containing entities as well as other, non-sulfur species that may be present in the repository. The effects of temperature, solution pH, and hydrogen pressure on the kinetics of the hydrogen electrode reaction (HER) on copper in borate buffer solution have been studied by means of steady-state polarization measurements, including electrochemical impedance spectroscopy (EIS). In order to obtain electrokinetic parameters, such as the exchange current density and the

  14. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  15. Nuclear relaxation and critical fluctuations in membranes containing cholesterol

    Science.gov (United States)

    McConnell, Harden

    2009-04-01

    Nuclear resonance frequencies in bilayer membranes depend on lipid composition. Our calculations describe the combined effects of composition fluctuations and diffusion on nuclear relaxation near a miscibility critical point. Both tracer and gradient diffusion are included. The calculations involve correlation functions and a correlation length ξ =ξ0T/(T -Tc), where T -Tc is temperature above the critical temperature and ξ0 is a parameter of molecular length. Several correlation functions are examined, each of which is related in some degree to the Ising model correlation function. These correlation functions are used in the calculation of transverse deuterium relaxation rates in magic angle spinning and quadrupole echo experiments. The calculations are compared with experiments that report maxima in deuterium and proton nuclear relaxation rates at the critical temperature [Veatch et al., Proc. Nat. Acad. Sci. U.S.A. 104, 17650 (2007)]. One Ising-model-related correlation function yields a maximum 1/T2 relaxation rate at the critical temperature for both magic angle spinning and quadrupole echo experiments. The calculated rates at the critical temperature are close to the experimental rates. The rate maxima involve relatively rapid tracer diffusion in a static composition gradient over distances of up to 10-100 nm.

  16. Models test on dynamic structure-structure interaction of nuclear power plant buildings

    International Nuclear Information System (INIS)

    Kitada, Y.; Hirotani, T.

    1999-01-01

    A reactor building of an NPP (nuclear power plant) is generally constructed closely adjacent to a turbine building and other buildings such as the auxiliary building, and in increasing numbers of NPPs, multiple plants are being planned and constructed closely on a single site. In these situations, adjacent buildings are considered to influence each other through the soil during earthquakes and to exhibit dynamic behaviour different from that of separate buildings, because those buildings in NPP are generally heavy and massive. The dynamic interaction between buildings during earthquake through the soil is termed here as 'dynamic cross interaction (DCI)'. In order to comprehend DCI appropriately, forced vibration tests and earthquake observation are needed using closely constructed building models. Standing on this background, Nuclear Power Engineering Corporation (NUPEC) had planned the project to investigate the DCI effect in 1993 after the preceding SSI (soil-structure interaction) investigation project, 'model tests on embedment effect of reactor building'. The project consists of field and laboratory tests. The field test is being carried out using three different building construction conditions, e.g. a single reactor building to be used for the comparison purposes as for a reference, two same reactor buildings used to evaluate pure DCI effects, and two different buildings, reactor and turbine building models to evaluate DCI effects under the actual plant conditions. Forced vibration tests and earthquake observations are planned in the field test. The laboratory test is planned to evaluate basic characteristics of the DCI effects using simple soil model made of silicon rubber and structure models made of aluminum. In this test, forced vibration tests and shaking table tests are planned. The project was started in April 1994 and will be completed in March 2002. This paper describes an outline and the summary of the current status of this project. (orig.)

  17. Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

    International Nuclear Information System (INIS)

    Xie Hui; Zhou Jie; He Yingchao

    1993-01-01

    Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

  18. Behaviour of concrete nuclear containment structures upto ultimate failure with special reference to MAPP-1 containment

    International Nuclear Information System (INIS)

    Appa Rao, T.V.S.R.

    1975-01-01

    Theoretical and experimental methods for investigating the behaviour of concrete secondary containment structures subjected to loads upto their ultimate failure have been discussed in the paper. Need for inelastic nonlinear analysis of containments has been emphasized. Different contitutive models of concrete that can be employed in the nonlinear analysis of concrete structures were briefly reviewed. Based on the experimental results obtained in a 1:12 scale model test conducted at the Structural Engineering Research (Regional) Centre, Madras, behaviour of the MAPP-1 containment to internal pressure loading upto its ultimate failure has been discussed. (author)

  19. Nuclear fission energy: new build, operation, fuel cycle and decommissioning in the international perspective

    Energy Technology Data Exchange (ETDEWEB)

    Niessen, Stefan [AREVA GmbH, Erlangen (Germany)

    2015-07-01

    Over 60 nuclear power reactors are in construction today and over 400 are connected to the grid. The presentation will show where. A nuclear new build project involves a team of several thousand people. Some pictures from ongoing new build projects will illustrate this. Using concrete examples from the AREVA group, the nuclear fuel cycle from uranium mines in Niger, Kazakhstan or Canada to chemical conversion, enrichment and fuel manufacturing will be explained. Also the recycling of used fuel and the fabrication of MOX fuel is addressed. The presentation closes with an overview on decommissioning and final storage projects.

  20. Risk ranking of LANL nuclear material storage containers for repackaging prioritization.

    Science.gov (United States)

    Smith, Paul H; Jordan, Hans; Hoffman, Jenifer A; Eller, P Gary; Balkey, Simon

    2007-05-01

    Safe handling and storage of nuclear material at U.S. Department of Energy facilities relies on the use of robust containers to prevent container breaches and subsequent worker contamination and uptake. The U.S. Department of Energy has no uniform requirements for packaging and storage of nuclear materials other than those declared excess and packaged to DOE-STD-3013-2000. This report describes a methodology for prioritizing a large inventory of nuclear material containers so that the highest risk containers are repackaged first. The methodology utilizes expert judgment to assign respirable fractions and reactivity factors to accountable levels of nuclear material at Los Alamos National Laboratory. A relative risk factor is assigned to each nuclear material container based on a calculated dose to a worker due to a failed container barrier and a calculated probability of container failure based on material reactivity and container age. This risk-based methodology is being applied at LANL to repackage the highest risk materials first and, thus, accelerate the reduction of risk to nuclear material handlers.

  1. Carolinas' Nuclear Cluster: building competency through collaboration

    Energy Technology Data Exchange (ETDEWEB)

    Little, J. [Carolinas' Nuclear Cluster (United States)

    2013-07-01

    This presentation discusses the Carolinas Nuclear Cluster that was built with collaboration amongst interested parties. The challenge facing the participants were availability of qualified & experienced workforce; retiring expertise; competition for resources within nuclear and other technology sectors; competition for skills and leadership; competing priorities in a changing environment such as slow growth in new nuclear in the U.S.; speedup in existing plant upgrades & retrofits and international project development. The established principles were collaboration amongst players, no competition, no borders, business driven focus on job creation, think as a global business, be willing to invest actively with money, talent, time and focus on results and not activities.

  2. Carolinas' Nuclear Cluster: building competency through collaboration

    International Nuclear Information System (INIS)

    Little, J.

    2013-01-01

    This presentation discusses the Carolinas Nuclear Cluster that was built with collaboration amongst interested parties. The challenge facing the participants were availability of qualified & experienced workforce; retiring expertise; competition for resources within nuclear and other technology sectors; competition for skills and leadership; competing priorities in a changing environment such as slow growth in new nuclear in the U.S.; speedup in existing plant upgrades & retrofits and international project development. The established principles were collaboration amongst players, no competition, no borders, business driven focus on job creation, think as a global business, be willing to invest actively with money, talent, time and focus on results and not activities.

  3. Building technology on construction site of nuclear power plant at Zaporozh'e

    Energy Technology Data Exchange (ETDEWEB)

    Dusek, R; Matyas, V [Vodni Stavby, Prague (Czechoslovakia)

    1981-12-01

    Basic data and technical and economic indexes are shown for a WWER 1000 nuclear power plant being built 120 km off Zaporozh'e (USSR). The schedule of construction and the choice of the means of mechanization used for building work are reported. Discussed are building machines used, the location of assembly cranes of the main unit, the design and the building technology of the reactor part, the engine house, deisel generator station, the special operations building, the use of concrete and steel units in the building, and the procurement of materials for the construction. The knowledge gained from the building of the power plant will be applied in the CSSR in the building of 1000 MW unit power plants.

  4. Nonlinear seismic behavior of a CANDU containment building subjected to near-field ground motions

    International Nuclear Information System (INIS)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2004-01-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near the fault. Near-fault ground motions are the ground motions that occur near an earthquake fault. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. In this study, nonlinear dynamic time-history analyses were performed to investigate the seismic behavior of a CANDU containment structure subjected to various earthquake ground motions including the near-field ground motions

  5. Nuclear magnetic resonance characterization of apple juice containing enzyme preparations

    International Nuclear Information System (INIS)

    Prestes, Rosilene A.; Almeida, Denise Milleo; Barison, Andersson; Pinheiro, Luis Antonio; Wosiacki, Gilvan

    2012-01-01

    In this work, 1 H nuclear magnetic resonance ( 1 H NMR) was employed to evaluate changes in apple juice in response to the addition of Panzym Yieldmash and Ultrazym AFP-L enzymatic complexes and compare it with premium apple juice. The juice was processed at different temperatures and concentrations of enzymatic complexes. The differences in the results were attributed mainly to the enzyme concentrations, since temperature did not cause any variation. A quantitative analysis indicated that the concentration of fructose increased while the concentrations of sucrose and glucose decreased in response to increasing concentrations of the enzymatic complexes. (author)

  6. An international data base of nuclear concrete containment ageing

    International Nuclear Information System (INIS)

    Seni, C.; Ianko, L.

    1994-01-01

    The ageing of nuclear structures is of special interest because of the extended service life expected of these structures, and the potential impact of their deterioration on safety and reliability. Although there are databases about concrete, they address properties in general, not performance. In 1992, the IAEA, in collaboration with AECL, set out to create a new database that would fill the gap. Functional ageing, i.e. deterioration of leak-tightness, was to be included, not just structural ageing, i.e. deterioration of load-bearing capacity. This paper outlines the project of creating the database

  7. Risk of containment sump plugging in EDF nuclear power plants

    International Nuclear Information System (INIS)

    2003-10-01

    The fuel of PWR type reactors in usual functioning is cooled by water circulating in the primary circuit, kept at a 155 bars pressure. In the case of break in the circuit, the reactor is stopped automatically. But it is necessary to evacuate the power that continues to escape from the fuel, because of the radioactivity of products created during the reactor operating. In this aim a system called system of security injection allows to send water in the reactor core when the vapor that is releases by the primary circuit opening is condensed in the containment by the spraying system in the containment that sprays water under the containment dome. The two systems of security and spraying are supplied by a tank but when it is empty they are supplied by water recovered in the bottom of containment in sumps. The two systems operate in closed circuit and allow to evacuate the residual power. The purpose of this work is to study the risk of filters clogging that are in the sumps. (N.C.)

  8. Building better optical model potentials for nuclear astrophysics applications

    International Nuclear Information System (INIS)

    Bauge, Eric; Dupuis, Marc

    2004-01-01

    In nuclear astrophysics, optical model potentials play an important role, both in the nucleosynthesis models, and in the interpretation of astrophysics related nuclear physics measurements. The challenge of nuclear astrophysics resides in the fact that it involves many nuclei far from the stability line, implying than very few (if any) experimental results are available for these nuclei. The answer to this challenge is a heavy reliance on microscopic optical models with solid microscopic physics foundations that can predict the relevant physical quantities with good accuracy. This use of microscopic information limits the likelihood of the model failing spectacularly (except if some essential physics was omitted in the modeling) when extrapolating away from the stability line, in opposition to phenomenological models which are only suited for interpolation between measured data points and not for extrapolating towards unexplored areas of the chart of the nuclides.We will show how these microscopic optical models are built, how they link to our present knowledge of nuclear structure, and how they affect predictions of nuclear astrophysics models and the interpretation of some key nuclear physics measurements for astrophysics

  9. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  10. Radiolysis effects on fuel corrosion within a failed nuclear waste container

    International Nuclear Information System (INIS)

    Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

    2003-01-01

    The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

  11. Problems of building social and production bases for the construction of the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Suchopar, J.

    1984-01-01

    A survey is given of capital costs of building parts and the concept is presented of the equipment of a building site. The social basis consists of accommodation units, cloakrooms, catering units, utility units, a health care centre and a shopping centre. The production base consists of a central concrete plant, cement plant, central armour unit, forming hall, storage area, buildings for pre-assembly, fleet of trucks and other machines, maintenance and repair shops. The network infrastructure and the mains connections are designed and the procedure of the build up of the site is described. The experience is summed up and recommendations made for future nuclear power plant projects. (E.S.)

  12. Nuclear-powered icebreaking container ship via the Northern Sea Route in its economical aspects

    International Nuclear Information System (INIS)

    Kondo, Koichi; Takamasa, Tomoji

    1996-01-01

    In recent years, major maritime nations such as Japan, Russia and Norway have been investigating the use of the Northern Sea Route (NSR) as a sea route between the Far East and Europe, linking the eastern and western parts of the Eurasian continent. In this study, as part of the examination of suitable merchant ships for the NSR, we make an economic comparison between diesel container ships taking the Suez Canal route and NSR nuclear-powered icebreaking container ship carrying the Marine Reactor X (MRX), which is currently being developed at the Japan Atomic Energy Research Institute. Compared to diesel container ships going via the Suez Canal, the first-year transportation cost of NSR nuclear-powered container ship after commissioning is 30-70% higher and the required freight rate (RFR) is 8-40% higher. If the nuclear reactor in nuclear-powered container ship, which is the reason for higher cost, were replaced by the cassette-type MRX, the reusability of the MRX would reduce this cost difference between nuclear-powered and diesel ships. The study also shows that in terms of the total cost including sales opportunity costs, NSR nuclear-powered container ship can compete sufficiently with diesel container ships on the Suez Canal route. (author)

  13. A complete analysis of a nuclear building to nuclear safety standards

    International Nuclear Information System (INIS)

    Bergeretto, G.; Giuliano, V.; Lazzeri, L.

    1975-01-01

    The nuclear standards impose on the designer the necessity of examining the loads, stresses and strains in a nuclear building even under extreme loading conditions, both due to plant malfunctions and environmental accidents. It is necessary then to generate, combine and examine a tremendous amount of data; really the lack of symmetry and general complication of the structures and the large number of loading combinations make an automatic analysis quite necessary. A largely automatized procedure is presented in view of solving the problem by a series of computer programs linked together as follows. After the seismic analysis has been performed by (SADE CODE) these data together with the data coming from thermal specifications, weight, accident descriptions etc. are fed into a finite element computer code (SAP4) for analysis. They are processed and combined by a computer code (COMBIN) according to the loading conditions (the usual list in Italy is given and briefly discussed), so that for each point (or each selected zone) under each loading condition the applied loads are listed. These data are fed to another computer code (DTP), which determines the amount of reinforcing bars necessary to accommodate the most severe of the loading conditions. The Aci 318/71 and Italian regulation procedures are followed; the characteristics of the program are briefly described and discussed. Some particular problems are discussed, e.g. the thermal stresses due to normal and accident conditions, the inelastic behavior of some frame elements (due to concrete cracking) is considered by means of an 'ad hoc' code. Typical examples are presented and the results are discussed showing a relatively large benefit in considering this inelastic effect

  14. Dynamic analysis and structural design of underground nuclear reactor containments

    International Nuclear Information System (INIS)

    Kierans, T.W.; Reddy, D.V.

    1975-01-01

    All concept options are assumed to be similar in design criteria for structural competence to contain radioactivity and fuel heat and meet the functional, servicing, protective and aesthetic requirements. The choice of underground siting should be based on criteria developed from the sequential consideration of load-causing phenomena, concept and site characteristics. From the criteria, loads for a particular concept and site are calculated and the design formulated. (orig./ORU) [de

  15. A Hydrogen Containment Process for Nuclear Thermal Engine Ground testing

    Science.gov (United States)

    Wang, Ten-See; Stewart, Eric; Canabal, Francisco

    2016-01-01

    The objective of this study is to propose a new total hydrogen containment process to enable the testing required for NTP engine development. This H2 removal process comprises of two unit operations: an oxygen-rich burner and a shell-and-tube type of heat exchanger. This new process is demonstrated by simulation of the steady state operation of the engine firing at nominal conditions.

  16. Method of detecting leakage in nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Koba, Akitoshi; Goto, Seiichiro.

    1974-01-01

    Object: To permit accurate and prompt detection of leakage of a radioactive substance. Structure: The rate of change of such factors as radiation dose, temperature and pressure in the containment vessel, and each detected rate of change is compared with a reference value. The running cycle of the condensed drain exhausting pump in a drain collecting tank within a predetermined period is detected, and it is also compared with a reference value. These comparisons determine the absence or presence of leakage. (Kamimura, M.)

  17. Chemical durability of simulated nuclear glasses containing water

    International Nuclear Information System (INIS)

    Li, H.; Tomozawa, M.

    1995-04-01

    The chemical durability of simulated nuclear waste glasses having different water contents was studied. Results from the product consistency test (PCT) showed that glass dissolution increased with water content in the glass. This trend was not observed during MCC-1 testing. This difference was attributed to the differences in reactions between glass and water. In the PCT, the glass network dissolution controlled the elemental releases, and water in the glass accelerated the reaction rate. On the other hand, alkali ion exchange with hydronium played an important role in the MCC-1. For the latter, the amount of water introduced into a leached layer from ion-exchange was found to be much greater than that of initially incorporated water in the glass. Hence, the initial water content has no effect on glass dissolution as measured by the MCC-1 test

  18. Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

    International Nuclear Information System (INIS)

    Kusserow, J.; Gurr, W.; Pflug, H.

    1985-05-01

    Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

  19. Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3

    International Nuclear Information System (INIS)

    Ulm, Franz-Josef

    2000-01-01

    OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3(NOTE: Part II A item 1 indicates ''PAPER'', but a report is attached electronically)

  20. Mesothelioma among employees with likely contact with in-place asbestos-containing building materials.

    Science.gov (United States)

    Anderson, H A; Hanrahan, L P; Schirmer, J; Higgins, D; Sarow, P

    1991-12-31

    The occurrence of mesothelioma is a sentinel event in occupational and environmental disease. A mesothelioma surveillance system was established utilizing existing computerized Wisconsin vital statistics data maintained since 1959 and a Cancer Reporting System (CRS) established in 1978. Review of the death certificate listing of usual occupation and industry from 487 mesothelioma deaths in Wisconsin from 1959 to 1989 led to the investigation of 41 persons with likely exposure to inplace asbestos-containing building materials (ACBM): 12 school teachers, 10 school maintenance employees, 7 public building maintenance workers, 5 private building maintenance workers, and 7 commercial and factory workers performing maintenance activities. For 10 (34%) of the 29 maintenance workers the only source of asbestos exposure identified was their maintenance work. For five (17%) histories indicated some prior employment in occupations and industries with probable asbestos exposures. Opportunities for indirect occupational exposure were identified for ten who had been employed in the residential construction industry. One maintenance worker was exposed to asbestos in the household and another had neighborhood exposure. For 9 (75%) of the school teachers, the only identifiable potential source of asbestos exposure was derived from in-place ACBM in schools. One teacher had spent a season in the merchant marine aboard an iron ore-hauling ship and 2 had worked in the residential construction industry. Two of the teachers were sisters, and in two instances, two teachers had taught in the same school facility. We conclude that individuals occupationally exposed to in-place ACBM are at risk for the subsequent development of mesothelioma.

  1. Annual energy analysis of concrete containing phase change materials for building envelopes

    International Nuclear Information System (INIS)

    Thiele, Alexander M.; Jamet, Astrid; Sant, Gaurav; Pilon, Laurent

    2015-01-01

    Highlights: • Adding PCM to concrete walls can significantly reduce the cooling needs of buildings. • Climate, season, and wall orientation strongly affect energy and cost savings. • The PCM melting temperature should be near the desired indoor temperature. • Benefits are maximum for outdoor temperature oscillating around set indoor temperature. • Adding PCM had little effect on heating energy needs and associated cost savings. - Abstract: This paper examines the annual energy and cost savings potential of adding microencapsulated phase change material to the exterior concrete walls of an average-sized single family home in California climate zones 3 (San Francisco, CA) and 9 (Los Angeles, CA). The annual energy and cost savings were larger for South- and West-facing walls than for other walls. They were also the largest when the phase change temperature was near the desired indoor temperature. The addition of microencapsulated phase change material to the building walls reduced the cooling load in summer substantially more than the heating load in winter. This was attributed to the cold winter temperatures resulting in nearly unidirectional heat flux on many days. The annual cooling load reduction in an average-sized single family home in San Francisco and in Los Angeles ranged from 85% to 100% and from 53% to 82%, respectively, for phase change material volume fraction ranging from 0.1 to 0.3. The corresponding annual electricity cost savings ranged from $36 to $42 in San Francisco and from $94 to $143 in Los Angeles. From an energy standpoint, the best climate for using building materials containing uniformly distributed microencapsulated phase change material would have outdoor temperature oscillations centered around the desired indoor temperature for the entire year

  2. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Allen, M.D.; Pilch, M.M.

    1994-01-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

  3. Clearance of building structures for conventional non-nuclear reuse

    International Nuclear Information System (INIS)

    Buss, K.; Boehringer, S.

    1998-01-01

    At the example of a fuel assembly plant the strategy of control measurements on building surfaces, which shall be conventionally reused after their clearance, is regarded. Based on the given clearance levels the used measuring methods, especially with regard of possibly covered or intruded uranium contamination, are shown. The possibility of using the in-situ-γ-spectroscopy is discussed. (orig.) [de

  4. Building control for nuclear materials R and D facility

    International Nuclear Information System (INIS)

    Hart, O.

    1979-01-01

    The new plutonium research and development facility at LASL was the first facility to be completed in the United States under the new environmental requirements. To insure that these new requirements are met, a redundant computer system is used to monitor and control the building. This paper describes the supervisory control and data acquisition system that was implemented to perform that function

  5. Safety analysis of the nuclear chemistry Building 151

    International Nuclear Information System (INIS)

    Kvam, D.

    1984-01-01

    This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables

  6. Hydrogen combustion study in the containment of Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.; Gonzalez Videla, E.

    1997-01-01

    In this paper the combustion of hydrogen was modeled and studied in the containment vessel of the Atucha I nuclear power station using the CONTAIN package. The hydrogen comes from the oxidation of metallic materials during the severe accidents proposed. The CONTAIN package is an integrated tool that analyzes the physical, chemical and radiation conditions that affect the containment structure of the radioactive materials unloaded from the primary system during a severe accident in the reactor. (author) [es

  7. Concrete containments in Swedish nuclear power plants. A review of construction and material

    International Nuclear Information System (INIS)

    Roth, Thomas; Silfwerbrand, Johan; Sundquist, Haakan

    2002-12-01

    attention. Current investigation shows that the documentation on the concrete containment structures of the Swedish nuclear power stations is fairly complete after the authors have obtained new information through a survey during 2001 and included these data in the report. The target group of this report are structural engineers and other people interested in knowing how the prestressed concrete containments in the Swedish nuclear power stations are designed, detailed and constructed. Uprising questions regarding the structural behaviour of the containment structures ought to be evaluated by using present material properties and not the data describing the used building materials at the design stage. The aim of this research project is to gain new knowledge on life span questions regarding prestressing steel in concrete structures, partly generally and partly with focus on Swedish nuclear power stations and Swedish bridges. The project covers both bonded and un bonded prestressing steel. This report describes the containment structures for all Swedish nuclear power stations. The information is both given in Chapters 5 through 16 and assembled in tables in Appendix A. The intention is that the documentation shall grow and be supplemented as soon as new information, either new data describing the containment structures or new measuring results, will be obtained or produced within current research project. Design and detailing of prestressed concrete structures are among others based on the knowledge of time-dependent material changes regarding concrete (creep and shrinkage) and prestressing steel (relaxation). The intention is that the following items will treated: general evaluation; testing of prestressing steel and concrete properties; assessment of the risk of a time-dependent increase of brittleness of the prestressing steel; comparisons with codes; modelling of steel relaxation; unidimensional modelling of prestressing losses; regard to elevated temperatures

  8. Experimental study of a laboratory concrete material representative of containment buildings: desorption isotherms and permeability determination

    International Nuclear Information System (INIS)

    Semete, P.; Fevrier, B.; Delorme, J.; Sanahuja, J.; Desgree, P.; Le Pape, Y.

    2015-01-01

    The isotherm sorption curve is a first order parameter for the calculations of concrete drying and/or creep using Finite Element Analysis. An experimental campaign was undertaken by EDF MMC in order to characterize the first desorption isotherm at room temperature of a laboratory material representative of concrete containment buildings. Long term drying tests were carried out on cement paste and on three samples geometries on concrete (with radial and axial one-dimensional drying on thin disks and multi-dimensional drying on Representative Elementary Volumes). The measurements results (porosity, densities and mass loss curves) are provided and the isotherms obtained for the four different configurations are compared. Several analyses of the results are proposed including the assessment of a criterion for the determination of the moisture content final balance (estimation of the asymptotic mass loss) and the back-analysis of equivalent permeability. (authors)

  9. Building a universal nuclear energy density functional (UNEDF)

    Energy Technology Data Exchange (ETDEWEB)

    Nazarewicz, Witold [Univ. of Tennessee, Knoxville, TN (United States)

    2012-07-01

    The long-term vision initiated with UNEDF is to arrive at a comprehensive, quantitative, and unified description of nuclei and their reactions, grounded in the fundamental interactions between the constituent nucleons. We seek to replace current phenomenological models of nuclear structure and reactions with a well-founded microscopic theory that delivers maximum predictive power with well-quantified uncertainties. Specifically, the mission of this project has been three-fold: First, to find an optimal energy density functional (EDF) using all our knowledge of the nucleonic Hamiltonian and basic nuclear properties. Second, to apply the EDF theory and its extensions to validate the functional using all the available relevant nuclear structure and reaction data. Third, to apply the validated theory to properties of interest that cannot be measured, in particular the properties needed for reaction theory.

  10. Analysis to the criticality the storage and containers to the Juragua Nuclear Power Plant Fuel

    International Nuclear Information System (INIS)

    Guerra Valdes, R.

    1998-01-01

    Presently analysis the criticality the warehouses and containers the nuclear fuels in Juragua nuclear power plant the property multiplicity determined in these system and it is verified that for the geometry and operation conditions defined in the design as well as in accidents situations, the arrangement the fuel stays subcritical with an appropriate margin

  11. New nuclear build and evolving radiological protection challenges

    International Nuclear Information System (INIS)

    Lazo, T.

    2010-01-01

    Many trends and indicators suggest that the use of nuclear power for generating electricity will increase, perhaps significantly, in the coming 10 to 20 years and beyond. Any such expansion will not take place in a static scientific or social context, but rather in the midst of ongoing changes in many relevant fields, radiological protection, radioactive waste management and nuclear safety to name a few. Regarding radiological protection, this evolution can be characterised in many different ways, but can conveniently be described as having scientific and socially driven aspects. These may well pose challenges to radiological protection (RP) policy, regulation and application in the future

  12. Long-term reliability evaluation of nuclear containments with tendon force degradation

    International Nuclear Information System (INIS)

    Kim, Sang-Hyo; Choi, Moon-Seock; Joung, Jung-Yeun; Kim, Kun-Soo

    2013-01-01

    Highlights: • A probabilistic model on long-term degradation of tendon force is developed. • By using the model, we performed reliability evaluation of nuclear containment. • The analysis is also performed for the case with the strict maintenance programme. • We showed how to satisfy the target safety in the containments facing life extension. - Abstract: The long-term reliability of nuclear containment is important for operating nuclear power plants. In particular, long-term reliability should be clarified when the service life of nuclear containment is being extended. This study focuses not only on determining the reliability of nuclear containment but also presenting the reliability improvement by strengthening the containment itself or by running a strict maintenance programme. The degradation characteristics of tendon force are estimated from the data recorded during in-service inspection of containments. A reliability analysis is conducted for a limit state of through-wall cracking, which is conservative, but most crucial limit state. The results of this analysis indicate that reliability is the lowest at 3/4 height of the containment wall. Therefore, this location is the most vulnerable for the specific limit state considered in this analysis. Furthermore, changes in structural reliability owing to an increase in the number of inspecting tendons are analysed for verifying the effect of the maintenance program's intensity on expected containment reliability. In the last part of this study, an example of obtaining target reliability of nuclear containment by strengthening its structural resistance is presented. A case study is conducted for exemplifying the effect of strengthening work on containment reliability, especially during extended service life

  13. Specific problems concerning aircraft impact on nuclear containment vessels

    International Nuclear Information System (INIS)

    Fuzier, J.P.; Cheyrezy, M.H.; Dufour, C.J.

    1977-01-01

    Due to the high population density, in Belgium PWR power plants are designed against aircraft impacts (BOEING 707 crashing at 360 km per hour and STARFIGHTER F 104 G crashing at 540 km per hour). A double wall is used for the containment shield. The lack of relevant data and specifications for such a loading on the non-prestressed external wall led the authors to determine the suitable safety criteria, the most appropriate materials to be used and the corresponding limit state design through dynamic and plastic analysis. (Auth.)

  14. The impact of team building and leadership development on nuclear plant performance

    International Nuclear Information System (INIS)

    Fiedler, P.B.; Long, R.L.; Childress, J.R.

    1988-01-01

    Within the nuclear utility industry, the pressures of complex technologies, increasing regulations, and critical public scrutiny create a working environment filled with numerous pressures. The difficult nature of the industry puts a premium on effective teamwork, interdepartmental cooperation, and communication skills. A well-conceived and implemented team building and leadership development program can substantially improve the operating performance of a nuclear plant. This paper describes one such implementation effort at GPU Nuclear Corporation and at the Oyster Creek nuclear generating station (OCNGS) over an 18-month period

  15. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  16. Nuclear plants: the more they build, the more you pay

    International Nuclear Information System (INIS)

    Lanoue, R.

    1976-01-01

    Taxpayers and consumers are shown to have taken on an expensive burden in pursuit of peaceful uses of nuclear energy when a better approach would be to develop smaller-scale alternative sources of energy and shift investment from energy consumption to energy conservation and efficiency. Economic and environmental pleasure and security are more likely to result from taking some of the suggested conservation measures in homes, businesses, and industries. Small-scale energy needs can be met by solar, wind, geothermal, and other sources on an indivudal basis, while larger-scale needs can be met with improved methods of burning and treating coal. A return to the free market will allow investments to be diverted from nuclear technology to other areas. Economic data illustrates the rising costs of nuclear fuels, technology, and plant construction, and stresses and economic liability of the breeder reactor. State and federal forums where citizens can intervene are listed for those wishing to take direct action. Directions are given for evaluating construction and operating cost estimates and for challenging the nuclear effort. 217 references

  17. Building A Universal Nuclear Energy Density Functional (UNEDF)

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Joe [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Furnstahl, Dick [The Ohio State Univ., Columbus, OH (United States); Horoi, Mihai [Central Michigan Univ., Mount Pleasant, MI (United States); Lusk, Rusty [Argonne National Lab. (ANL), Argonne, IL (United States); Nazarewicz, Witek [Univ. of Tennessee, Knoxville, TN (United States); Ng, Esmond [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Thompson, Ian [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Vary, James [Iowa State Univ., Ames, IA (United States)

    2012-09-30

    During the period of Dec. 1 2006 - Jun. 30, 2012, the UNEDF collaboration carried out a comprehensive study of all nuclei, based on the most accurate knowledge of the strong nuclear interaction, the most reliable theoretical approaches, the most advanced algorithms, and extensive computational resources, with a view towards scaling to the petaflop platforms and beyond. The long-term vision initiated with UNEDF is to arrive at a comprehensive, quantitative, and unified description of nuclei and their reactions, grounded in the fundamental interactions between the constituent nucleons. We seek to replace current phenomenological models of nuclear structure and reactions with a well-founded microscopic theory that delivers maximum predictive power with well-quantified uncertainties. Specifically, the mission of this project has been three-fold: first, to find an optimal energy density functional (EDF) using all our knowledge of the nucleonic Hamiltonian and basic nuclear properties; second, to apply the EDF theory and its extensions to validate the functional using all the available relevant nuclear structure and reaction data; third, to apply the validated theory to properties of interest that cannot be measured, in particular the properties needed for reaction theory. The main physics areas of UNEDF, defined at the beginning of the project, were: ab initio structure; ab initio functionals; DFT applications; DFT extensions; reactions.

  18. Seismic analysis, evaluation and upgrade design for a nuclear facility exhaust stack building

    International Nuclear Information System (INIS)

    Malik, L.E.; Kabir, A.F.

    1991-01-01

    This paper reports on an exhaust stack building of a nuclear reactor facility with complex structural configuration that has been analyzed and evaluated for seismic forces. This building was built in the 1950's and had not been designed to resist seismic forces. A very rigorous analysis and evaluation program was implemented to minimize the costly retrofits required to upgrade the building to resist high seismic forces. The seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, base mat flexibility and the influence of the nearby reactor building have been considered in the seismic analyses. The rigorous analyses and evaluation enabled limited upgrades to qualify the stack building for the seismic forces

  19. The Nuclear Power Landscape in the Post-Fukushima Era: Thoughts on the Current and Future Effects on Nuclear New-Build

    International Nuclear Information System (INIS)

    Tyobeka, B.

    2012-01-01

    The nuclear landscape looked very promising before the Fukushima Daiichi accident. In the past five years before the accident, so-called Nuclear Rennainsance looked to be in full swing, with many countries beginning to factor nuclear energy as part of their electricity generation mix. At some point, 43 IAEA Members States confirmed their interest in launching new nuclear power programmes. Whilst only two of these nuclear new-comers have already chosen the reactor designs they would deploy in their new build, it is commonly accepted that the so-called Generation III and Generation III+ would mostly be the designs of the choice for new nuclear build. This presentation seeks to examine the current status of plans for nuclear build after Fukushima, looking into technology and safety issues that would influence the final policy decisions in new nuclear build programmes.(author).

  20. Thirty years of measured prestress at Swedish nuclear reactor containments

    International Nuclear Information System (INIS)

    Anderson, Patrick

    2005-01-01

    The main function of the reactor containment, i.e. to ensure tightness at a major internal accident, depends directly on the prestressing system. To secure that the prestress level is sufficient, the tendon force has been measured during the whole time of operation. The general results from these measurements show that the loss of prestress 30 years after tensioning is between 5 and 10%. This is much lower loss than predicted initially at the design stage. More advanced and today commonly used models for predicting prestress loss show better agreement with the results. The main reasons for the relatively low loss are assumed to be: (1) the confirmed slow drying process of the concrete and (2) the high concrete age at the initial tensioning. The results also indicate that the temperature has a major influence on the loss of prestress

  1. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  2. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  3. Building Foundations for Nuclear Security Enterprise Analysis Utilizing Nuclear Weapon Data

    Energy Technology Data Exchange (ETDEWEB)

    Josserand, Terry Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis; Young, Leone [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis; Chamberlin, Edwin Phillip [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Security Enterprise and Cost Analysis

    2017-09-01

    The Nuclear Security Enterprise, managed by the National Nuclear Security Administration - a semiautonomous agency within the Department of Energy - has been associated with numerous assessments with respect to the estimating, management capabilities, and practices pertaining to nuclear weapon modernization efforts. This report identifies challenges in estimating and analyzing the Nuclear Security Enterprise through an analysis of analogous timeframe conditions utilizing two types of nuclear weapon data - (1) a measure of effort and (2) a function of time. The analysis of analogous timeframe conditions that utilizes only two types of nuclear weapon data yields four summary observations that estimators and analysts of the Nuclear Security Enterprise will find useful.

  4. Building roof with conical holes containing PCM to reduce the cooling load: Numerical study

    International Nuclear Information System (INIS)

    Alawadhi, Esam M.; Alqallaf, Hashem J.

    2011-01-01

    Highlights: → We investigated a roof structure with cone frustum holes containing PCM. → Different types of PCM and geometries of the PCM are presented. → The results indicate that the heat flux at the indoor space is reduced by 39%. - Abstract: The thermal effectiveness of a building's roof with phase change material (PCM) is presented in this paper. The considered model consists of a concrete slab with vertical cone frustum holes filled with PCM. The objective of incorporating the PCM into the roof structure is to utilize its high latent heat of fusion to reduce the heat gain during the energy demanded peak hours, by absorbing the incoming energy through the melting process in the roof before it reaches the indoor space. The thermal effectiveness of the proposed roof-PCM system is determined by comparing the heat flux at the indoor surface to a roof without the PCM during typical working hours. A parametric study is conducted to assess the effects of the cone frustum geometry, and the kind of PCM used. The n-Eicosane shows the best performance among the examined PCMs, and the conical geometry of the PCM container is the best in term of thermal effectiveness. The results indicate that the heat flux at the indoor surface of the roof can be reduced up to 39% for a certain type of PCM and geometry of PCM cone frustum holes.

  5. Education and Training Networks as a Tool for Nuclear Security Human Resource Development and Capacity Building

    International Nuclear Information System (INIS)

    Nikonov, D.

    2014-01-01

    Human Resource Development for Capacity Building for Nuclear Security: • Comprehensive Training Programme Objective: To raise awareness, to fill gaps between the actual performance of personnel and the required competencies and skills and, to build-up qualified instructors/trainers. • Promoting Nuclear Security Education Objective: To support the development of teaching material, faculty expertise and preparedness, and the promotion of nuclear security education in collaboration with the academic and scientific community. Ultimate Goal: To develop capabilities for supporting sustainable implementation of the international legal instruments and IAEA guidelines for nuclear security worldwide, and to foster nuclear security culture. Education priorities for the future: • Incorporate feedback from the first pilot program into future academic activities in nuclear security; • Based on feedback from pilot program: • Revise the NSS12 guidance document; • Update educational materials and textbooks. • Support INSEN members, which consider launching MSc programs at their institutions; • Continue promoting nuclear security education as part of existing degree programs (through certificate or concentration options); • Support the use of new forms of teaching and learning in nuclear security education: • Online e-learning degree programmes and modules; • Learning by experience; • Problem-oriented learning tailored to nuclear security functions

  6. Mitigating Capability Analysis during LOCA for Korean Standard Nuclear Power Plants in Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, Soo Yong; Kim, D. H.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    The objective of this paper is to establish Containment spray operational technical bases for the typical Korean Standard Nuclear Power plants (Ulchin units 3 and 4) by modeling the plant, and analyzing a loss of coolant accident (LOCA) using the MAAP code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately by computer codes. Furthermore, it is important to attain the capability to analyze a advanced nuclear reactor design for a severe accident prevention and mitigation.

  7. Probabilistic analysis of the efficiency of the damping devices against nuclear fuel container falling

    Science.gov (United States)

    Králik, Juraj

    2017-07-01

    The paper presents the probabilistic and sensitivity analysis of the efficiency of the damping devices cover of nuclear power plant under impact of the container of nuclear fuel of type TK C30 drop. The finite element idealization of nuclear power plant structure is used in space. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall. The experimental results of the shock-damper basic element behavior under impact loads are presented. The Newmark integration method is used for solution of the dynamic equations. The sensitivity and probabilistic analysis of damping devices was realized in the AntHILL and ANSYS software.

  8. Blasting jobs on the building site of the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Voda, J.; Podel, R.

    1984-01-01

    The problems are discussed of the preparation and implementation on the Temelin nuclear power plant building site of blasting, the volume of drilling and the choice of the drill hammer - all based on experience gained during the construction of the Mochovce nuclear power plant. The amount of explosives used on the Temelin site will be 1400 t. The use of mechanical charging may shorten the preparation of the site by 20 to 30%. Explosive emulsion slurries are being developed from home raw materials whose application will reduce the volume of drilling by 15%. The method of controlled breaking secures adequate quality of peripheral walls and bottom chink but special explosives will have to be used. Seismic effects are discussed of blasting on dwellings, agricultural and industrial buildings in the vicinity of the site, on-site buildings, underground mains and special structures of the nuclear power plant. (E.S.)

  9. Commissioning of the steel containment and its related components of the Loviisa II. nuclear power plant

    International Nuclear Information System (INIS)

    Tuominen, J.; Pietikaeinen, L.; Kutramoinen, H.

    1982-01-01

    The outer concrete wall of the containment building serves as a protective system for the components in side. It contains the hermetically sealed steel pressure vessel for retaining the release of radioactive contamination in an accident situation. During a loss-of-coolant accident the pressure is reduced in two steps. The various testing procedures of the containment locks, their main-tenance and repair, the pressure and tightness tests of the steel containment and the preliminary operational tests of the other components of the containment system has been presented. (R.P.)

  10. Preparation for and building of V-1 nuclear power plant

    International Nuclear Information System (INIS)

    Smatlak, S.; Spirko, M.

    1981-01-01

    Some findings and problems encountered during the construction of the V-1 nuclear power plant at Jaslovske Bohunice are discussed from the point of view of the investor, mainly problems associated with project preparation and relations between suppliers and the customer. The progress of construction is described and the dates are shown of the main stages of construction and tests. The construction proceeded according to a net diagram and was controlled by the construction management and a coordination group headed by the investor. An international start-up commission was established for the start-up stage. The fields are outlined of the participation and co-operation of Soviet organizations during project preparation, supplies, assembly, start-up, and fuel supplies for the V-1 nuclear power plant. (B.S.)

  11. Nuclear island buildings layout collaborative design platform development and application

    International Nuclear Information System (INIS)

    Chen Li; Huang Wenqiang

    2014-01-01

    Based on characteristics of nuclear island layout design, the large number files and complicated interface, for realizing collaborative design and fine management goal, Establish collaborative design platform, which includes the design task module, 3D design module, project management module. These three modules can package design input files, realize synchronous design and real-time track design drawings state, timely feedback between design, procurement, construction site. There is no design task delay due to tracking and has realized fine management of design. (authors)

  12. Building

    OpenAIRE

    Seavy, Ryan

    2014-01-01

    Building for concrete is temporary. The building of wood and steel stands against the concrete to give form and then gives way, leaving a trace of its existence behind. Concrete is not a building material. One does not build with concrete. One builds for concrete. MARCH

  13. Hydrogen deflagration simulations under typical containment conditions for nuclear safety

    Energy Technology Data Exchange (ETDEWEB)

    Yanez, J., E-mail: jorge.yanez@kit.edu [Institute for Energy and Nuclear Technology, Karlsruhe Institute of Technology, Kaiserstrasse 12, 76131 Karlsruhe (Germany); Kotchourko, A.; Lelyakin, A. [Institute for Energy and Nuclear Technology, Karlsruhe Institute of Technology, Kaiserstrasse 12, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Lean H{sub 2}-air combustion experiments highly relevant to typical NPP simulated. Black-Right-Pointing-Pointer Analyzed effect of temperature, concentration of H{sub 2}, and steam concentration. Black-Right-Pointing-Pointer Similar conditions and H{sub 2} concentration yielded different combustion regimes. Black-Right-Pointing-Pointer Flame instabilities (FIs) were the effect driving divergences. Black-Right-Pointing-Pointer Model developed for acoustic FI in simulations. Agreement experiments obtained. - Abstract: This paper presents the modeling of low-concentration hydrogen deflagrations performed with the recently developed KYLCOM model specially created to perform calculations in large scale domains. Three experiments carried out in THAI facility (performed in the frames of international OECD THAI experimental program) were selected to be analyzed. The tests allow studying lean mixture hydrogen combustion at normal ambient, elevated temperature and superheated and saturated conditions. The experimental conditions considered together with the facility size and shape grant a high relevance degree to the typical NPP containment conditions. The results of the simulations were thoroughly compared with the experimental data, and the comparison was supplemented by the analysis of the combustion regimes taking place in the considered tests. Results of the analysis demonstrated that despite the comparatively small difference in mixture properties, three different combustion regimes can be definitely identified. The simulations of one of the cases required of the modeling of the acoustic-parametric instability which was carefully undertaken.

  14. A study on the hydrogen distributions in a containment for nuclear plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kweon Ha; Kim, Ju Youn; Bae, Kyung Hyo [The Korea Maritime Univ., Busan (Korea, Republic of)

    2012-10-15

    Hydrogen explosion has been considered as one of the major issues since Fukushima nuclear accident. The cause of the explosion has not been discovered, but it is clear that the explosion strongly depends on hydrogen distributions in a containment. In this study hydrogen distributions are calculated and analyzed in the containment of APR 1400(Advanced Power Reactor 1400)

  15. Modelling and simulation of containment on full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zou Tingyun

    1996-01-01

    A multi-node containment thermal-hydraulic model has been developed and adapted in Full Scope Simulator for Qinshan 300 MW Nuclear Power Unit with good realtime simulation effects. Containment pressure for LBLOCA calculated by the model is well agreed with those of CONTEMPT-4/MOD3

  16. General requirements for concrete containment structures for CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1993-07-01

    This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts

  17. Define rules for the exporter and importer of minerals or ores containing nuclear elements

    International Nuclear Information System (INIS)

    1969-01-01

    The present resolution establishes regulations for the exporter of minerals or ores containing associated nuclear elements, and for the importer of chemical compounds of technical purity grade, containing a quantity of fissile of fertile materials equal to the existent in the exported material

  18. Use of the radiochemical method for identification of seized nuclear material - container

    International Nuclear Information System (INIS)

    Rosskopfova, O.; Matel, L.; Rajec, P.; Dobias, M.

    2003-01-01

    Analysis of the nuclear material of the seized illicit trafficking container was performed in the laboratory LARCHA. The container was professionally taken apart in the hot cell of HUMA-LAB APEKO. The compact core of container with mass of 36.00 kg was separated and analysed in the LARCHA laboratory. The results of alpha spectrometry proved that the core of container was made of depleted uranium. A suspicious container was seized, that was believed to be used for transportation of radioactive and nuclear materials. Container, according to Slovak Law, was transported to a Civil defence and police started with expertise analysis. On the basis of previous results, it was assumed that the core of the container was made of a nuclear material. The container was cut inside a hot cell to separate parts at HUMA-Lab Kosice and individual parts were prepared for analyses. From shape and type of the detained container it was assumed that the core of the container was made of depleted uranium and suspected cut pieces were sent to the accredited laboratory LARCHA Bratislava for isotope analysis. Samples from different parts of the container were analysed by using HPGe gamma. The results of gamma spectrometry proved the presence of Co-60, Cs-137 and U-235 with values 11.8 Bq, 5.6 Bq a 3.8 Bq respectively. On the basis of measured results it was calculated mass content (%) of uranium isotopes in samples. It was concluded that: - container is composed of different parts, which are made from steel, lead and uranium; - the wipe tests proved the presence of radionuclides of Cs-137 and Co-60; - the core of the container is made of depleted uranium; - chosen methods, which were used proved that are suitable for analysis and identification of nuclear materials. (authors)

  19. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  20. Total organizational team building at the V.C. Summer Nuclear Station

    International Nuclear Information System (INIS)

    Carswell, R.W.; Parks, J.W.

    1988-01-01

    The V.C. Summer station of South Carolina Electric and Gas Company began a total organizational team building program in 1985 as one of several methods by which to make overall operational improvements. The team model selected was developed by Carswell ampersand Associated, Inc. The model has been used successfully in a number of nonnuclear applications. This was its first usage in a nuclear power plant. Total organizational team building has had positive results at the V.C. Summer station. The data suggest that including all employees in decision making can assist with increasing efficiency, morale, and safety in a nuclear power plant

  1. Hyperpolarized Xenon Nuclear Magnetic Resonance (NMR of Building Stone Materials

    Directory of Open Access Journals (Sweden)

    Michele Mauri

    2012-09-01

    Full Text Available We have investigated several building stone materials, including minerals and rocks, using continuous flow hyperpolarized xenon (CF-HP NMR spectroscopy to probe the surface composition and porosity. Chemical shift and line width values are consistent with petrographic information. Rare upfield shifts were measured and attributed to the presence of transition metal cations on the surface. The evolution of freshly cleaved rocks exposed to the atmosphere was also characterized. The CF-HP 129Xe NMR technique is non-destructive and it could complement currently used techniques, like porosimetry and microscopy, providing additional information on the chemical nature of the rock surface and its evolution.

  2. Development of Capacity Building Training Programs for Nuclear R and D Personnel

    International Nuclear Information System (INIS)

    Lee, Eui Jin; Nam, Youngmi; Hwang, Hyeseon; Jang, Eunsook; Song, Eun Ju

    2016-01-01

    The Nuclear Training and Education Center of the Korea Atomic Energy Research Institute has been operating technical training courses on nuclear engineering, engineering mathematics, management leadership training, out sourced practical training, legal education, etc. Strengthening nuclear R and D capacity is essential for the long-term mission and goals of the institute. Therefore, it requires a comprehensive training program to strengthen the unique capability of the institute that reflects diversity and differentiation. In this regard, the capacity building training program has developed on a modular basis, and the developed training program should be tailored to operate according to the institute needs. The capacity building training program for nuclear R and D personnel was developed to reflect the technology strengths of the institute. The developed training program will be developed into a leading branded education of the institute in the future

  3. International Conference on Human Resource Development for Nuclear Power Programmes: Building and Sustaining Capacity. Presentations

    International Nuclear Information System (INIS)

    2014-01-01

    The objectives of the conference are to: • Review developments in the global status of HRD since the 2010 international conference; • Emphasize the role of human resources and capacity building programmes at the national and organizational level for achieving safe, secure and sustainable nuclear power programmes; • Discuss the importance of building competence in nuclear safety and security; • Provide a forum for information exchange on national, as well as international, policies and practices; • Share key elements and best practices related to the experience of Member States that are introducing, operating or expanding nuclear power programmes; • Highlight the practices and issues regarding HRD at the organizational and national level; • Highlight education and training programmes and practices; • Emphasize the role of nuclear knowledge management for knowledge transfer and HRD; and • Elaborate on the role and scope of various knowledge networks

  4. Development of Capacity Building Training Programs for Nuclear R and D Personnel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eui Jin; Nam, Youngmi; Hwang, Hyeseon; Jang, Eunsook; Song, Eun Ju [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Nuclear Training and Education Center of the Korea Atomic Energy Research Institute has been operating technical training courses on nuclear engineering, engineering mathematics, management leadership training, out sourced practical training, legal education, etc. Strengthening nuclear R and D capacity is essential for the long-term mission and goals of the institute. Therefore, it requires a comprehensive training program to strengthen the unique capability of the institute that reflects diversity and differentiation. In this regard, the capacity building training program has developed on a modular basis, and the developed training program should be tailored to operate according to the institute needs. The capacity building training program for nuclear R and D personnel was developed to reflect the technology strengths of the institute. The developed training program will be developed into a leading branded education of the institute in the future.

  5. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bosland, L. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France)], E-mail: loic.bosland@irsn.fr; Funke, F. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany); Girault, N. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France); Langrock, G. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany)

    2008-12-15

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two

  6. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Bosland, L.; Funke, F.; Girault, N.; Langrock, G.

    2008-01-01

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two different kinetic models

  7. The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

    International Nuclear Information System (INIS)

    Kondo, K.; Takamasa, T.

    2000-01-01

    The design concept of the cassette-type-reactor MRX (Marine Reactor X), being under development in Japan for the nuclear ice-breaker container ship is described. The MRX reactor is the monoblock water-cooled and moderated reactor with passive cooling system of natural circulation. It is shown that application of the reactor being under consideration gives an opportunity to decrease greatly the difference in prices for similar nuclear and diesel ships. Economic estimations for applicability of the nuclear ice-breaker container ship with the MRX reactor in Arctics for transportation of standard containers TEU from Europe to Far East as compared with transportation of the same containers by diesel ships via Suets Canal are made [ru

  8. Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

    International Nuclear Information System (INIS)

    Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

    1999-01-01

    A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

  9. Growth of capital investments for nuclear power plaut building in the USA

    International Nuclear Information System (INIS)

    Gitlevich, A.D.

    1985-01-01

    Data concerning the growth of capital investments for building NPPs in the USA are presented. Experts are explaining a considerable growth of NPP building cost by changes in standards and regulations approved earlier, long terms of licensing procedures and building period delay caused by the necessity to meet increased safety and environment requirements. Nonetheless the American experts suppose that in the eighties and subsequent +ears the main types of power plants in the USA will remain the nuclear Uower plants along with the thermal ones, although according to predictions the capital investments for the NPP planned to be put in operation in 1992 will exceed those for the thermal ones by 61%

  10. Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

    International Nuclear Information System (INIS)

    Hagan, R.; Gladson, J.

    1997-01-01

    The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

  11. Structural design of the turbine building of Angra Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Varella, L.N.; Reis, F.J.C.; Jurkiewicz, W.J.

    1978-01-01

    The Turbine Building of the Angra Nuclear Power Plant, Unit 1, and particularly its structure and structural design are described. The Turbine Building, as far as its structure is concerned, deviates from the standard structure of any turbine building due to the fact that huge ducts are provided in the foundation mat as to accomodate the circulating water system. This aspect and the fact that the building is founded upon a very deep strata of compacted and controlled fill, makes out of the building structure 'a concrete ship floating in the sea of sand', and by the same reason presents by itself an interesting structure, worth to be known to all engineers involved in design of power plants. This pape, suplemented by a few slides shown during presentation of the paper at the conference, covers the subject mainly from the designers' point of view. (Author)

  12. Integration of the security systems in the architectural design of nuclear and important buildings in Egypt

    International Nuclear Information System (INIS)

    Algohary, S.

    2007-01-01

    The new and emerging threats to buildings and infrastructure which are faced by todays engineering design and facility management community in Egypt demand new approaches and solutions that are innovative and increasingly based on risk management principles. In the wake of the damage of Taba hotel in south Sinai (2004) and Sharm El-Sheik hotels in Egypt (July, 2005), there was a growing awareness of public vulnerability to terrorist attacks. This awareness leads to increase the expectations form and responsibilities of the architects, engineers and construction professionals This study reviews and assesses different types of threats to nuclear and important buildings. It identifies also the architectural design, vulnerability and risk management that can enhance security. It also introduces a new approach for integration of architectural design and security in nuclear and important buildings in Egypt. The results shows that escalating threats and risks to important buildings and infrastructures change the role of planners, architects, engineers and builders by increasing the focus on the importance of applying viable security principles to the building designs. Architects in Egypt can assume an important role in improving the life-safety features of important buildings by increasing and integrating new security principles and approaches to improve the security and performance of the buildings against man made disasters

  13. Method of decommissioning nuclear reactor building by utilizing sea water buyoancy

    International Nuclear Information System (INIS)

    Iwashima, Sumio; Ogoshi, Shigeru; Kobari, Shin-ichi.

    1989-01-01

    Upon dismantling nuclear reactor buildings, peripheral yards are excavated and channels leading to sea shore are formed. Since the outer walls of the reactor buildings are made of iron-reinforced concretes, the opening poritons are grouted with concretes to attain a tightly such closed structure that radioactive wastes, etc. in the inside are not flown out upon reactor discommisioning. Peripheral buildings at relatively low level of radiation contaminations are dismantled and withdrawn. The fundations of the nuclear reactor buildings were dug out and jacked to separate base rocks and the reactor buildings. Then, sea water is introduced into the water channels to entirely float up the buildings. A water gate is disposed in the water channel on the side of sea shore to control the level of sea water. The buildings are moved and guided to the sea shore and towed to a site optimum as a permanent storage area and then burried in that place. The operation period for the discommissioning work can greatly be shortened and the radiation dose and the amount of the wastes can be reduced. (T.M.)

  14. A New European COST Network 'NORM4Building' (TU1301) for the Reuse of NORM Containing Residues in Building Materials

    International Nuclear Information System (INIS)

    Schroeyers, W.; Schreurs, S.

    2014-01-01

    The new COST action was initiated on the 1st of January 2014 and runs for four years. COST is supported by the EU RTD Framework Program. In the presentation more information on how to participate in the network will be provided. In the presentation the new approach and new initiatives of the NORM4BUILDING network, that has its first meeting here in the DEAD SEA Hotel on the 12-13/02/2014, will be introduced. The NORM4Building materials network will be an open network of researchers. An Advisory Board consisting mainly from NORM processing and construction industries and relevant associations and regulators are invited to work in collaboration with the scientists that will populate the various working groups and the management committee of the new COST action

  15. Pressure test at the reactor building of the Embalse Nuclear Power Plant (CNE)

    International Nuclear Information System (INIS)

    Coutsiers, E.E.; Perrino, J.; Moreno, C.; Batistic, J.A.; Lolis, R.R.; Aviles, A.

    1991-01-01

    Upon request by the Licensing Authority, the reactor building (RB) in a nuclear power plant must be submitted to pressure tests. One of these tests is to be performed before startup and, then, a test must be carried out every 5 years in operation. The pre-operational tests took place in August 1981, under two values of relative pressure: 1.266 kg/cm 2 and 0.422 kg/cm 2 . Operational tests must only be made at the lower pressure and their objective is to verify that the loss speed remains within the range indicated in the corresponding technical specification. The first operational test was performed in August 1989. The personnel of the CNE took care of the preparation of the Work Plan, of aligning the various systems contained in the RB, of pressurization, of monitoring localized tightedness, of depressurization and of the general and quality control of the test. The measurements were carried out by the CISME (Center of Metrology Research and Service) of the National Institute of Industrial Technology (INTI) , which did also supply the necesary instruments and the data collection system. There is also a description of the work performed before the test, of the calculation method used for assessing the loss rate, of the test sequencies and of the results obtained. (Author) [es

  16. Building capacity through leadership development programmes in nuclear industry

    International Nuclear Information System (INIS)

    Afonin, A.; Terentyeva, T.

    2014-01-01

    «Tournament of Emerging Professionals» (TEMP) – an effective tool for selecting and developing talented students. Aim: to select high-potential graduates from key Rosatom universities through a multi-stage competition and employ them at Rosatom enterprises (NPPs, fuel plants, research institutes etc). Key objectives: 1. To provide opportunities for graduates and students to apply theoretical; technical knowledge while working on real case-studies and projects provided by Rosatom enterprises; 2. To popularize nuclear industry and to motivate high-potential young people with technical education to work in Rosatom and its enterprises; 3. To involve subject matter experts and TOP-managers of Rosatom into interactions with high-potential students and graduates

  17. Intake of radioactivity by inhalation, in buildings after nuclear incidents

    International Nuclear Information System (INIS)

    Brenk, H.D.

    1987-01-01

    Risk studies so far assumed that in case of a nuclear incident, the exposure of persons through inhalation proceeds in free air by direct intake of contaminated ambient air. In reality, however, about 80 p.c. of the population is indoors, and there is a considerable delay of room air contamination in comparison with free air contamination. The radionuclides are deposited on indoor surfaces (such as walls, windows, furniture, etc.), and in air-conditioning systems on the filters. This contamination is somewhat regularly removed through cleaning or filter exchange, which reduces the contamination of room air, and then also the radiation exposure by inhalation. The reducing effect can be enhanced by such simple measures as closing doors and windows in time, and switching the air-conditioning system to recirculating regime, followed by enhanced fresh air supply at a later time. (orig./HP) [de

  18. Containment response analysis for the PSA (Probabilistic Safety Assessment) of the CAREM-25 nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.

    1997-01-01

    This work is part of the probabilistic safety assessment actually under development for the CAREM-25 nuclear power station, and departs from the accident sequences already obtained and quantified by the Event Trees/Fault Trees techniques. At first, the potential containment failure modes for nuclear stations are listed, based on the experience. Then, the CAREM-25 design peculiarities are analyzed, on their possible influence on the containment behavior during severe accidents. Then plan damage states are defined. Furthermore, containment damage states are also defined, and containment event trees are built for each plant damage state. Those sequences considered representative from the annual probability (those which exceed or probability of IE-09 per year, are used to quantify the combinations of plant damage states/containment damage states, based on the estimation of a vulnerability matrix. (author) [es

  19. Care of personnel on the building site of the Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Kurial, P.

    1984-01-01

    The accommodation is described of workers on the building site of the Dukovany nuclear power plant. The quality is appraised of accommodation, catering and refreshments. There is a health care unit on site and 15 beds are reserved at the Trebic hospital for emergency cases. Trade union and youth organizations look after sports and cultural activities. (E.S.)

  20. Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

    International Nuclear Information System (INIS)

    Duncan, D.W.

    1995-01-01

    This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

  1. Cost, Schedule and Risk Management, The Building Blocks of a U.S. Nuclear Project

    International Nuclear Information System (INIS)

    Redding, John

    2002-01-01

    The most important consideration in the decision to build a new nuclear plant is the capital cost. Right? Yes and no. Yes, the capital cost accounts for 80% of the generation cost of a new plant. No, because there are other equally important considerations. (author)

  2. Seismic simulation analysis of nuclear reactor building by soil-building interaction model

    International Nuclear Information System (INIS)

    Muto, K.; Kobayashi, T.; Motohashi, S.; Kusano, N.; Mizuno, N.; Sugiyama, N.

    1981-01-01

    Seismic simulation analysis were performed for evaluating soil-structure interaction effects by an analytical approach using a 'Lattice Model' developed by the authors. The purpose of this paper is to check the adequacy of this procedure for analyzing soil-structure interaction by means of comparing computed results with recorded ones. The 'Lattice Model' approach employs a lumped mass interactive model, in which not only the structure but also the underlying and/or surrounding soil are modeled as descretized elements. The analytical model used for this study extends about 310 m in the horizontal direction and about 103 m in depth. The reactor building is modeled as three shearing-bending sticks (outer wall, inner wall and shield wall) and the underlying and surrounding soil are divided into four shearing sticks (column directly beneath the reactor building, adjacent, near and distant columns). A corresponding input base motion for the 'Lattice Model' was determined by a deconvolution analysis using a recorded motion at elevation -18.5 m in the free-field. The results of this simulation analysis were shown to be in reasonably good agreement with the recorded ones in the forms of the distribution of ground motions and structural responses, acceleration time histories and related response spectra. These results showed that the 'Lattice Model' approach was an appropriate one to estimate the soil-structure interaction effects. (orig./HP)

  3. Risks attached to container- and bunker-storage of nuclear waste

    International Nuclear Information System (INIS)

    Jager, D. de

    1987-12-01

    The results are presented of a literature study into the risks attached to the two dry-storage options selected by the Dutch Central Organization For Radioactive Waste (COVRA): the container- and the bunker-storage for irradiated nuclear-fuel elements and nuclear waste. Since the COVRA does not make it clear how these concepts should have to be realized, the experiences abroad with dry interim-storage are considered. In particular the Castor-container-storage and the bunker storage proposed in the committee MINSK (Possibilities of Interim-storage in the Netherlands of Irradiated nuclear-fuel elements and Nuclear waste) are studied further in depth. The committee MINSK has performed a study into the technical realizability of various interim-storage facilities, among which a storage in bunkers. (author). 75 refs.; 14 figs.; 16 tabs

  4. Development of deterioration models and tests of structural materials for nuclear containment structures(III)

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Byung Hwan [Seoul National University, Seoul (Korea)

    2002-03-01

    The nuclear containment structures are very important infrastructures which require much cost for construction and maintenance. If these structures lose their functions and do not ensure their safety, great losses of human lives and properties will result. Therefore, the nuclear containment structures should secure appropriate safety and functions during these service lives. The nuclear concrete structures start to experience deterioration due to severe environmental condition, even though the concrete structures exhibit generally superior durability. It is, therefore, necessary to take appropriate actions at each stage of planning, design and construction to secure safety and functionability. Thorough examination of deterioration mechanism and comprehensive tests have been conducted to explore the durability characteristics of nuclear concrete structures. 88 refs., 70 figs., 12 tabs. (Author)

  5. Burning rates of hydrogen-air mixtures in containment buildings and the consequent pressure transients

    International Nuclear Information System (INIS)

    Tennankore, K.N.; Kumar, R.K.; Razzaghi, M.

    1987-01-01

    One-dimensional flame models are often used to predict the pressure transients caused by hydrogen combustion in containments during postulated severe accidents. In the absence of data, these models account for prevailing flame acceleration mechanisms, such as initial turbulence, venting and obstacle-induced turbulence, by using arbitrarily large burning velocities that are much higher than laminar burning velocities. Using an intermediate-scale test facility at the Whiteshell Nuclear Research Establishment we have obtained necessary data on the effects of flame acceleration mechanisms, to estimate the safety margin in the buring velocities used in the models. So far, data have been analyzed, with a one-dimensional model, to determine effective burning velocities and burning-rate enhancement factors. The results of the analyses indicate that the effect of initial turbulence on the burning rate can be bounded only if the effect of flame-generated turbulence is included. The effect of venting can be accounted for by using two burning velocities, one for the pre-vent duration and a second increased value during the vented-combustion stage. The enhancement factors due to these two mechanisms, for the different conditions analyzed, varied up to 5.4, and the effective burning velocities varied up to 8.4 m/s

  6. Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

    International Nuclear Information System (INIS)

    De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

    2005-01-01

    Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

  7. Calculations of hydrogen detonations in nuclear containments by the random choice method

    International Nuclear Information System (INIS)

    Delichatsios, M.A.; Genadry, M.B.

    1983-01-01

    Computer codes were developed for the prediction of pressure histories at different points of a nuclear containment wall due to postulated internal hydrogen detonations. These pressure histories are required to assess the structural response of a nuclear containment to hydrogen detonations. The compressible flow equations including detonation, which was treated as a sharp fluid discontinuity, were solved by the random choice method which reproduces maximum pressures and discontinuities sharply. The computer codes were validated by calculating pressure profiles and maximum wall pressures for plane and spherical geometries and comparing the results with exact analytic solutions. The two-dimensional axisymmetric program was used to calculate wall pressure histories in an actual nuclear containment. The numerical results for wall pressures are presented in a dimensionless form, which allows their use for different combinations of hydrogen concentration, and initial conditions. (orig.)

  8. Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

    International Nuclear Information System (INIS)

    Lin, Y.C.; Xie, Y.J.; Wang, X.H.

    2004-01-01

    Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software

  9. Effects of non-uniform embedments on earthquake responses of nuclear reactor building

    International Nuclear Information System (INIS)

    Koyanagi, Y.; Okamoto, S.; Yoshida, K.; Inove, H.

    1989-01-01

    The nuclear reactor buildings have the portion embedded in soil. In the seismic design of such structures, it is essential to consider the effects of the embedment on the earthquake response. Most studies on these effects, however, assume the uniform embedment, i.e. the depth of the embedment is constant, which is convenient for the design and analysis. The behavior of the earthquake response considering the three-dimensional aspects of non-uniform embedment has not been made clear yet. In this paper, the authors evaluate the effects of the non-uniform embedment in an inclined ground surface on the earthquake response of a nuclear reactor building as illustrated. A typical PWR type reactor building is chosen as an analysis structure model. Four different types of embedment are set up for the comparison study. The three-dimensional analysis is carried out considering the geometry of embedment

  10. Effects of different SSI parameters on the floor response spectra of a nuclear reactor building

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.; Maryak, M.E.

    1991-01-01

    The effects of several critical soil-structure interaction (SSI) parameters on the floor response spectra (FRS) of a typical nuclear reactor building have been examined. These parameters are computation of soil impedance functions using different approaches, scattering effects (reductions in ground motion due to embedment and rigidity of building foundation) and strain dependency of soil dynamic properties. This paper reports that the significant conclusions of the study, which are applicable to a deeply embedded very rigid nuclear reactor building, are as follows: FRS generated without considering scattering effects are highly conservative; differences between FRS, generated considering strain-dependency of soil dynamic properties, and those generated suing low-strain values, are not significant; and the lumped-parameter approach of SSI calculations, which only uses a single value of soil shear modulus in impedance calculations, may not be able to properly compute the soil impedances for a soil deposit with irregularly varying properties with depth

  11. The development of the nuclear physics in Latvia II. The building of the Research Nuclear Reactor IRT

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2004-01-01

    Nuclear research reactor IRT of the Academy of Sciences was built near Riga in Salaspils. IRT is pool aqueous - aqueous reactor with nuclear fuel U-235 contained elements, located in the core at a depth of ∼ 7 m under distilled water. Ten horizontal and 10-15 vertical experimental channels are employed in experimental research with the use of neutron fluxes. For the research with gamma rays is constructed radiation loop facility with liquid In-Ga-SN solid solution as intensive gamma-ray sources. Main activities of IRT are to conduct research in nuclear spectroscopy, neutron activation analysis, neutron diffraction and radiation physics, chemistry and biology. (authors)

  12. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Kitano, Takehito

    1997-01-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  13. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  14. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1989-01-01

    This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

  15. Description of the CONTAIN input model for the Dodewaard nuclear power plant

    International Nuclear Information System (INIS)

    Velema, E.J.

    1992-02-01

    This report describes the ECN standard CONTAIN input model for the Dodewaard Nuclear Power Plant (NPP) that has been developed by ECN. This standard input model will serve as a basis for analyses of the phenomena which may occur inside the Dodewaard containment in the event of a postulated severe accident. Boundary conditions for specific containment analyses can easily be implemented in the input model. as a result ECN will be able to respond quickly on requests for analyses from the utilities of the authorities. The report also includes brief descriptions of the Dodewaard NPP and the CONTAIN computer program. (author). 7 refs.; 5 figs.; 3 tabs

  16. Program to develop analytical tools for environmental and safety assessment of nuclear material shipping container systems

    International Nuclear Information System (INIS)

    Butler, T.A.

    1978-11-01

    This paper describes a program for developing analytical techniques to evaluate the response of nuclear material shipping containers to severe accidents. Both lumped-mass and finite element techniques are employed to predict shipping container and shipping container-carrier response to impact. The general impact problem is computationally expensive because of its nonlinear, three-dimensional nature. This expense is minimized by using approximate models to parametrically identify critical cases before more exact analyses are performed. The computer codes developed for solving the problem are being experimentally substantiated with test data from full-scale and scale-model container drop tests. 6 figures, 1 table

  17. Space-sharing Strategy for Building Dynamic Container Yard Storage Considering Uncertainty on Number of Incoming Containers

    Directory of Open Access Journals (Sweden)

    Siti Nurminarsih

    2017-12-01

    Full Text Available The implementation of AEC (Asian Economic Community leads to demand increase at ports in Indonesia. This increasing demand is followed by operational efficiency at the port as well. In fact, ports in Indonesia have an average dwell time for 5 days. One of the causes of this long dwell time is container transfer inefficiencies during loading process. In this research, we will focus in yard management and increasing land utilization. To increase the land utilization, we develop the space-sharing concept to reduce the initial space needed for a given workload. Since the ships arrivals are also varying, this concept will be possible. We develop the concept by adding uncertainty on number of incoming containers at operational level as the determining factor in planning a shared yard area. We propose a simulation approach to evaluate strategies in making yard template.

  18. Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

    Directory of Open Access Journals (Sweden)

    Králik Juraj

    2014-12-01

    Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

  19. Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

    International Nuclear Information System (INIS)

    Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

    1980-01-01

    The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

  20. Simulation and Experimental Validation of Electromagnetic Signatures for Monitoring of Nuclear Material Storage Containers

    International Nuclear Information System (INIS)

    Aker, Pamela M.; Bunch, Kyle J.; Jones, Anthony M.

    2013-01-01

    Previous research at the Pacific Northwest National Laboratory (PNNL) has demonstrated that the low frequency electromagnetic (EM) response of a sealed metallic container interrogated with an encircling coil is a strong function of its contents and can be used to form a distinct signature which can confirm the presence of specific components without revealing hidden geometry or classified design information. Finite element simulations have recently been performed to further investigate this response for a variety of configurations composed of an encircling coil and a typical nuclear material storage container. Excellent agreement was obtained between simulated and measured impedance signatures of electrically conducting spheres placed inside an AT-400R nuclear container. Simulations were used to determine the effects of excitation frequency and the geometry of the encircling coil, nuclear container, and internal contents. The results show that it is possible to use electromagnetic models to evaluate the application of the EM signature technique to proposed versions of nuclear weapons containers which can accommodate restrictions imposed by international arms control and treaty verification legislation