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Sample records for npp seismic design

  1. Seismic analysis and design of NPP structures

    International Nuclear Information System (INIS)

    de Carvalho Santos, S.H.; da Silva, R.E.

    1989-01-01

    Numerical methods for static and dynamic analysis of structures, as well as for the design of individual structural elements under the applied loads are under continuous development, being very sophisticated methods nowadays available for the engineering practice. Nevertheless, this sophistication will be useless if some important aspects necessary to assure full compatability between analysis and design are disregarded. Some of these aspects are discussed herein. This paper presents an integrated approach for the seismic analysis and design of NPP structures: the development of models for the seismic analysis, the distribution of the global seismic forces among the seismic-resistant elements and the criteria for the design of the individual elements for combined static and dynamic forces are the main topics to be discussed herein. The proposed methodology is illustrated. Some examples taken from the project practice are presented for illustration the exposed concepts

  2. Russian regulatory approaches to seismic design and seismic analysis of NPP piping

    International Nuclear Information System (INIS)

    Kaliberda, Y.V.

    2003-01-01

    The paper presents an overview of Russian regulatory approaches to seismic design and seismic analysis of NPP piping. The paper is focused on categorization and seismic analysis of nuclear power plant items (piping, equipment, supports, valves, but not building structures). The paper outlines the current seismic recommendations, corresponding methods with the examples of calculation models. The paper considers calculation results of the mechanisms of dynamic behavior and the problems of developing a rational and economical approaches to seismic design and seismic protection. (author)

  3. The regulatory requirements, design bases, researches and assessments in the field of Ukrainian NPP's seismic safety

    International Nuclear Information System (INIS)

    Mykolaychuk, O.; Mayboroda, O.; Krytskyy, V.; Karnaukhov, O.

    2001-01-01

    State Nuclear Regulatory Authority of Ukraine (SNRA) pays large attention to problem of nuclear installations seismic stability. As a result the seismic design regulatory guides is revised, additional seismic researches of NPP sites are conducted, seismic reassessment of NPP designs were begun. The experts involved address all seismic related factors under close contact with the staff of NPP, design institutes and research organizations. This document takes stock on the situation and the research programs. (author)

  4. Seismic site evaluation practice and seismic design guide for NPP in Continent of China

    Energy Technology Data Exchange (ETDEWEB)

    Yuxian, Hu [State Seismological Bureau, Beijing, BJ (China). Inst. of Geophysics

    1997-03-01

    Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

  5. Seismic site evaluation practice and seismic design guide for NPP in Continent of China

    International Nuclear Information System (INIS)

    Hu Yuxian

    1997-01-01

    Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

  6. The seismic reassessment Mochovce NPP

    International Nuclear Information System (INIS)

    Baumeister, P.

    2004-01-01

    The design of Mochovce NPP was based on the Novo-Voronez type WWER-440/213 reactor - twin units. Seismic characteristic of this region is characterized by very low activity. Mochovce NPP site is located on the rock soil with volcanic layer (andesit). Seismic reassessment of Mochovce NPP was done in two steps: deterministic approach up to commissioning confirmed value Horizontal Peak Ground Acceleration HPGA=0.1 g and activities after commissioning as a consequence of the IAEA mission indicate higher hazard values. (author)

  7. Seismic characterization of the NPP Krsko site

    International Nuclear Information System (INIS)

    Obreza, J.

    2000-01-01

    The goal of NPP Krsko PSA Project Update was the inclusion of plant changes (i.e. configuration/operational related) through the period January 1, 1993 till the OUTAGE99 (April 1999) into the integrated Internal/External Level 1/Level 2 NPP Krsko PSA RISK SPECTRUM model. NPP Krsko is located on seismotectonic plate. Highest earthquake was recorded in 1917 with magnitude 5.8 at a distance of 7-9 km. Site (founded) on Pliocene sediments which are as deep as several hundred meters. No surface faulting at the Krsko site has been observed and thus it is not to be expected. NPP Krsko is equipped with seismic instrumentation, which allows it to complete OBE (SSE). The seismic PSA successfully showed high seismic margin at Krsko plant. NPP Krsko seismic design is based on US regulations and standards

  8. Seismic response analysis and upgrading design of pump houses of Kozloduy NPP units 5 and 6

    International Nuclear Information System (INIS)

    Jordanov, M.; Marinov, M.; Krutzik, N.

    2001-01-01

    The main objective of the presented project was to perform a feasibility study for seismic/structural evaluation of the safety related structures at Kozloduy NPP Units 5 and 6 for the new site seismicity and determine if they satisfy current international safety standards. The evaluation of the Pump House 3 (PH3) building is addressed in this paper, which was carried out by applying appropriate modeling techniques combined with failure mode and seismic margin analyses. The scope of the work defined was to present the required enhancement of the seismic capacity of the Pump House structures.(author)

  9. Calculation of NPP pipeline seismic stability

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Ambriashvili, Yu.K.; Kaliberda, I.V.

    1982-01-01

    A simplified design procedure of seismic pipeline stability of NPP at WWER reactor is described. The simplified design procedure envisages during the selection and arrangement of pipeline saddle and hydraulic shock absorbers use of method of introduction of resilient mountings of very high rigidity into the calculated scheme of the pipeline and performance of calculations with step-by-step method. It is concluded that the application of the design procedure considered permits to determine strains due to seismic loads, to analyze stressed state in pipeline elements and supporting power of pipe-line saddle with provision for seismic loads to plan measures on seismic protection

  10. Seismic safety review mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP. Sofia, Bulgaria, 19-23 October 1992

    International Nuclear Information System (INIS)

    Ma, D.; Prato, C.; Godoy, A.

    1992-10-01

    A seismic Safety Review Mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP was performed in Sofia from 19-23 October 1992. The objectives of the mission were to assist the Bulgarian authorities in: the evaluation of the floor response spectra of the main buildings of units 1-4 at Kozloduy NPP, calculated for the new defined seismic parameters at site (Review Level Earthquake - RLE); the evaluation of the remedial and strengthening measures proposed for the seismic upgrading of the pump house and diesel generator buildings to the new defined RLE. This mission completed the scope of previous IAEA mission - BUL/9/012-18b - (see Report 3262) performed from 3-7 August 1992, with regard to tasks which were not evaluated at that time because they had not been finished. 2 tabs

  11. Investigation on seismic characteristic in Muria Peninsula to asses the NPP seismic design

    International Nuclear Information System (INIS)

    Kusnowo, A.

    1997-01-01

    A feasibility study on first nuclear power plant was conducted during 4,5 years started on 22 November 1991. This study consists of two parts. First, the non site study, Second part is the site and environmental study. The scope of non site studies are economic financing, technical and safety aspect as well as fuel cycle and waste management aspect. The site and environmental studied consist of site conditions and qualification, seismological, environmental condition as well as social economic and cultural impact. In the first step of site study (step 1), the result come up to the three candidates named Ujung Lemahabang, Ujung Watu and Ujung Grenggengan. Further study on geology, topography, oceanography, geophysics, hydrology, seismology, vulcanology, man induced event, etc was done on those three candidates (named as step 2). The results come up with Ujung Lemahabang as the best candidates. It is important to know basic, characteristic of seismicity of nuclear power plant sitting region for seismic hazard assessment this was done as step 3. This paper describe the results of step 3. (J.P.N.)

  12. Investigation on seismic characteristic in Muria Peninsula to asses the NPP seismic design

    Energy Technology Data Exchange (ETDEWEB)

    Kusnowo, A [National Atomic Energy Agency, Jakarta (Indonesia)

    1997-03-01

    A feasibility study on first nuclear power plant was conducted during 4,5 years started on 22 November 1991. This study consists of two parts. First, the non site study, Second part is the site and environmental study. The scope of non site studies are economic financing, technical and safety aspect as well as fuel cycle and waste management aspect. The site and environmental studied consist of site conditions and qualification, seismological, environmental condition as well as social economic and cultural impact. In the first step of site study (step 1), the result come up to the three candidates named Ujung Lemahabang, Ujung Watu and Ujung Grenggengan. Further study on geology, topography, oceanography, geophysics, hydrology, seismology, vulcanology, man induced event, etc was done on those three candidates (named as step 2). The results come up with Ujung Lemahabang as the best candidates. It is important to know basic, characteristic of seismicity of nuclear power plant sitting region for seismic hazard assessment this was done as step 3. This paper describe the results of step 3. (J.P.N.)

  13. Seismic evaluation and upgrading design of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Jordanov, M.J.; Stoyanov, G.S.; Geshanov, I.H.; Kirilov, K.P.; Schuetz, W.

    2003-01-01

    This paper presents results obtained during the study of overhead roads between Reactor Building (RB) of WWER-1000 MW NPP and possible measures for their seismic upgrade. The main objective of this project is to evaluate the behavior of overhead roads under site-specific seismic loading and to determine whether this structure satisfies current international safety regulations, followed by development of upgrading concepts. Overhead roads are pre-cast RC structure, which can be divided to separate substructures. They comprise of pedestrian gallery and pipeline box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated for Review Level Earthquake (RLE) as seismic category II structures. As seismic input motion is RLE, free field response spectra anchored to 0.2 g PGA are used with 0.5 scaling factor. Soil-Structure Interaction effects are taken into account through equivalent soil springs with frequency adjusted stiffness. In order to meet the objective of the project a technical design specification is developed for conformance with International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The separate steps comprising the overall fulfillment of project's major objectives may be summarized as follows: study of all available data for initial design and as built conditions, creation of 3-D detailed finite element models for as-built structure, determination of dynamic characteristics, evaluation of adequacy of initial design under new seismic loading (calculation of D/C ratios for structural members and connections, evaluation of embedment lengths for embedded parts and rebars, deformation evaluation, stability checks), development of upgrading concepts for enhancement, verification of capability of upgraded structure

  14. Seismic safety of building structures of NPP Kozloduy III

    International Nuclear Information System (INIS)

    Varbanov, G.I.; Kostov, M.K.; Stefanov, D.D.; Kaneva, A.D.

    2005-01-01

    In the proposed paper is presented a general summary of the analyses carried out to evaluate the dynamic behavior and to assess the seismic safety of some safety related building structures of NPP Kozloduy. The design seismic loads for the site of Kozloduy NPP has been reevaluated and increased during and after the construction of investigated Units 5 and 6. Deterministic and probabilistic approaches are applied to assess the seismic vulnerability of the investigated structures, taking into account the newly defined seismic excitations. The presented results show sufficient seismic safety for the studied critical structures and good efficiency of the seismic upgrading. The applicability of the investigated structures at sites with some higher seismic activities is discussed. The presented study is dealing mainly with the civil structures of the Reactor building, Turbine hall, Diesel Generator Station and Water Intake Structure. (authors)

  15. Estimation of dynamic loading on a design of the NPP caused by seismic influences

    International Nuclear Information System (INIS)

    Proskuryakov, Konstantin

    2011-01-01

    Methods and algorithms of calculations of quality factor of a stream of the coolant are developed. Quantitative estimations of a range of frequency of vibration - acoustical resonance between the coolant flowing through the reactor core and fuel assembly vibration in the NPP with WWER-1000 are provided. The design procedure of quality factor of a stream of the coolant and a band - width in advanced light water reactor is developed. The experimental substantiation of sharp increase of intensity of vibrations at occurrence of vibration - acoustical resonance is received. The reasons of abnormal growth of level of vibrations are identified at stationary modes of cold - ops, hot - ops of the equipment of reactor installations with WWER-1000. It is showed that for prevention of vibration - acoustical resonance of the coolant and fuel assembly it is necessary and sufficient to deduce own frequency of fuel assembly vibrations from band - width limits. The technique of designing of cartograms of a reactor core with indication of quantity and location of fuel assemblies with high level of vibration is worked out. (author)

  16. Seismic safety programme at NPP Paks. Propositions for coordinated international activity in seismic safety of the WWER-440 V-213

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper presents the Paks NPP seismic safety program, highlighting the specifics of the WWER-440/213 type in operation, and the results of work obtained so far. It covers the following scope: establishment of the seismic safety program (original seismic design, current requirements, principles and structure of the seismic safety program); implementation of the seismic safety program (assessing the seismic hazard of the site, development of the new concept of seismic safety for the NPP, assessing the seismic resistance of the building and the technology); realization of the seismic safety of higher level (technical solutions, drawings, realization); ideas and propositions for coordinated international activity

  17. Seismic qualification of civil engineering structures - Temelin NPP

    International Nuclear Information System (INIS)

    Schererova, K.; Holub, I.; Stepan, J.; Maly, J.

    2004-01-01

    Basic information is presented about the input data and methodology used for evaluation of Temelin NPP civil structures. The existing conditions as listed in POSAR report for the two reactor units are considered. The original design of the power plant assumed a lower level of locality seismic hazard, as followed from seismological surveys that where then available. Later the seismic assessment was updated while fully respecting IAEA recommendations and using a minimum value of acceleration in the horizontal direction PGAHOR = 0.1 g at free field level for SL-2. In relation to the new seismic project, new qualification of the structures, components and systems classed as seismic resistance category 1 was carried out. Since the Czech Republic has no specific technical standards for seismic resistance evaluation of nuclear power plants, a detailed methodology was elaborated, comprising principles of seismic resistance evaluation based on IAEA guides and on common practice in countries with advanced nuclear power engineering. (P.A.)

  18. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  19. Seismic re-evaluation process in Medzamor-2 NPP

    International Nuclear Information System (INIS)

    Zadoyan, P.

    2000-01-01

    Seismic re-evaluation process for Medzamor-2 NPP describes the following topics: program implementation status; re-evaluation program structure; regulatory procedure and review plan; current tasks and practice; and regulatory assessment and research programs

  20. Floor response spectra of WWER-1000, NPP Kozloduy generated from local seismic excitation

    International Nuclear Information System (INIS)

    Bojadziev, Z.; Kostov, M.

    1996-01-01

    The seismic review level characteristics for the Kozloduy NPP site were set to 0.2 g and a respective free field acceleration response spectra were derived after a profound site conformation project. Accordingly a separate investigation is recommended for local seismic excitation. The goals of the analyses are: to define the seismic motion characteristics from local seismic sources; to perform structural analyses and in-structure spectra generation for local seismic excitation; and to compare the forces (spectra) from local events with those generated as seismic design review basis

  1. Preliminary evaluation of the seismic hazard at Cernavoda NPP site

    International Nuclear Information System (INIS)

    Mingiuc, C.; Serban, V.; Androne, M.

    2001-01-01

    The probabilistic seismic hazard analysis (PSHA) is a methodology by which one evaluates the probability of exceeding different parameters of the ground motions (the maximum ground acceleration - PGA and the ground response spectrum - SA) as effect of the seismic action, on a given site at a future time moment. Due to the large uncertainties in the geological, geophysical, seismological input data, as well as, in the models utilised, various interpretation schemes are applied in the PSHA analyses. This interpretation schemes lead to opinion discrepancies among specialists which finally lead to disagreements in estimating the values of the seismic design for a given site. In order to re-evaluate the methodology and to improve the PSHA result stability, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) and Electric Power Research Institute (EPRI) sponsored a project for defining methodological guides of performing PSHA analyses. The project was implemented by a panel of 7 experts, the Senior Seismic Hazard Analysis Committee - SSHAC. This paper presents a preliminary evaluation of the seismic hazard for the Cernavoda NPP site by application of the methodology mentioned, by taking into account the possible sources which could affect the site (the Vrancea focus, Galati - Tulcea fault, Sabla - Dulovo fault and local earthquakes)

  2. Seismic safety review mission for the follow-up of the seismic upgrading of Kozloduy NPP (Units 1-4). Sofia, Bulgaria, 16-20 November 1992

    International Nuclear Information System (INIS)

    David, M.; Shibata, H.; Stevenson, J.D.; Godoy, A.; Gurpinar, A.

    1992-11-01

    A Seismic Safety Review Mission for the follow-up of the design and implementation of the seismic upgrading of Kozloduy NPP was performed in Sofia from 16-20 November 1992. This mission continued the second task of the follow-up activities of the design and implementation of the seismic upgrading (Phases 1 and 2), which is being carried out in Units 1 and 2 of the NPP. Thus the objectives of the mission was to assist the Bulgarian authorities in the technical evaluation of the design tasks defined for Phases 1 and 2 item HB of WANO 6 Month Programme, as follows: anchorage upgrades of low seismic capacity components; list of seismic safety related systems and components; detailed walkdown to assess seismic capacity of components and define priorities for the upgrading; determination of seismic structural capacity of pump house, diesel generator building and turbine building and design of required upgrades; liquefaction potential evaluation. Tabs

  3. Russian standards and design practice of ensuring NPP reliability under severe external loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Birbraer, A N [St. Petersburg Research and Design Institute Atomenergoproject, St. Petersburt (Russian Federation)

    1993-07-01

    Russian Standards and design practice of ensuring NPP reliability under severe external loading conditions are described. The main attention is paid to the seismic design requirements. Explosions, aircraft impact, and tornado are briefly examined too (author)

  4. Russian standards and design practice of ensuring NPP reliability under severe external loading conditions

    International Nuclear Information System (INIS)

    Birbraer, A.N.

    1993-01-01

    Russian Standards and design practice of ensuring NPP reliability under severe external loading conditions are described. The main attention is paid to the seismic design requirements. Explosions, aircraft impact, and tornado are briefly examined too (author)

  5. Determination of the NPP Cernavoda reactor building seismic response

    International Nuclear Information System (INIS)

    Krutzik, N.J.; Rotaru, I.; Bobei, M.; Mingiuc, C.; Serban, V.

    1997-01-01

    Seismic input for systems and equipment installed in buildings depends on: - the seismic movement in free field on site; - the building movement in the soil; - the building deflection. The percentage of the 3 movements for the system and equipment input, depends on the position of the systems and equipment inside the building as well on the type of the foundation soil. The type of the foundation soil is important because if it is stiff it transfers a lot of energy to the building, energy which amplify the movement of the building on the top. If the foundation soil is soft, it accommodates the overall movement of the building in the soil, amplifying the movement to lower levels and the building response is attenuated if a resonance phenomenon between the whole building movement and the seismic excitation does not exist. This input is given with the design floor response spectra (FRS), in the logarithmic scale and seismic anchor movement (SAM). The design floor response spectra for NPP Cernavoda U1 Nuclear Building were determined in several stages starting with simple models (STICK type) without twisting movement and ending with detailed 3-dimensional models. From the point of view of dynamic behavior, the Reactor Building can be considered to be made up of 4 sub-structures: the containment building, internal structures containing separate elements such as the reactor vault, the fuel transfer structure and itself. Each sub-structure has its own movement (some of the structures present also some local effects) which combines with the overall movement of the building in the soil and the seismic excitation produce the total movement so that the response spectrum for each point of the sub-structure is specific. One should note that for structures which also show the twisting effect, the selection of the points on the floor, for the determination on the response spectra, is an engineering decision so that the response should be relevant for the equipment installed on the

  6. Constructive approaches to the space NPP designing

    International Nuclear Information System (INIS)

    Eremin, A.G.; Korobkov, L.S.; Matveev, A.V.; Trukhanov, Yu.L.; Pyshko, A.P.

    2000-01-01

    An example of designing a space NPP intended for power supply of telecommunication satellite is considered. It is shown that the designing approach based on the introduction of a leading criterion and dividing the design problems in two independent groups (reactor with radiation shield and equipment module) permits to develop the optimal design of a space NPP [ru

  7. The enhancement of Ignalina NPP in design and operational safety

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1999-01-01

    Enhancement of Ignalina NPP design include: core design improvements; fuel channel integrity (multiple pressure tube rupture); improvements of shutdown systems; improvements of instrumentation and control devices; containment strength and tightness; design basis accident analysis; improvements of safety and support systems; seismic safety enhancement; Year 2000 project; cracks in pipes. Enhancement of operational safety includes: quality assurance; configuration management; safety management and safety culture; emergency operating procedures; training and full scope simulator; in-service inspection; fire protection and ageing monitoring and management

  8. Seismic re-evaluation of Kozloduy NPP criteria, methodology, implementation

    International Nuclear Information System (INIS)

    Kostov, M.

    2003-01-01

    The paper describes some features of the methodology applied for seismic upgrading of civil structures at the site of the Kozloduy NPP. The essence of the methodology is the use of as-build data, realistic damping and inelastic reduction factors. As an example of seismic upgrading the analyses of units 3 and 4 are presented. The analyses are showing that for effective seismic upgrading detailed investigations are needed in order to understand the significant response modes of the structures. In the presented case this is the rotation of the attached flexible structures to the stiff reactor building. Based on this an upgrading approach is applied to increase the seismic resistance for the predominant motion. The second significant approach applied is the strengthening of the prefabricated element joints. Although it is very simple it allows use of the available element capacity. (author)

  9. Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2005-03-15

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

  10. Seismic analysis for safety related structures of 900MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  11. Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

    International Nuclear Information System (INIS)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2005-03-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments

  12. Seismic hazard analysis of the NPP Kozloduy site

    International Nuclear Information System (INIS)

    Petrovski, D.; Stamatovska, S.; Arsovski, M.; Hadzievski, D.; Sokerova, D.; Solakov, D.; Vaptzarov, I.; Satchanski, S.

    1993-01-01

    The principal objective of this study is to define the seismic hazard for the NPP Kozloduy site. Seismic hazard is by rule defined by the probability distribution function of the peak value of the chosen ground motion parameter in a defined time interval. The overall study methodology consists of reviewing the existing geological, seismological and tectonic information to formulate this information into a mathematical model of seismic activity of the region and using this assess earthquake ground motion in terms of probability. Detailed regional and local seismological investigations have been performed. Regional investigations encompass the area within a radius of 320 km from the NPP Kozloduy site. The results of these investigations include all seismological parameters that are necessary for determination of the mathematical model of the seismicity of the region needed for the seismic hazard analysis. Regional geological and neotectonic investigations were also performed for the wider area including almost the whole territory of Bulgaria, a large part of Serbia, part of Macedonia and almost the whole south part of Romania

  13. Application of the SASSI soil structure interaction method to CANDU 6 NPP seismic analysis

    International Nuclear Information System (INIS)

    Ricciuti, R.A.; Elgohary, M.; Usmani, S.A.

    1996-01-01

    The standard CANDU 6 NPP has been conservatively qualified for a Design Basis Earthquake (DBE) peak horizontal ground acceleration of 0.2 g. Currently there are potential opportunities for siting the CANDU 6 at higher seismicity sites. In order to be able to extend the use of a standardized design for sites with higher seismicity than the standard plant, various design options, including the use of the SASSI Soil Structure Interaction (SSI) analysis method, are being evaluated. This paper presents the results of a study to assess the potential benefits from utilization of the SASSI computer program and the use of more realistic damping ratios for the structures

  14. 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.)

  15. 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.)

  16. Parametric Study on Ultimate Failure Criteria of Elbow Piping Components in Seismically Isolated NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Ki, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    It is well known that the interface pipes between isolated and non-isolated structures will become the most critical in the seismically isolated NPPs. Therefore, seismic performance of such interface pipes should be evaluated comprehensively especially in terms of the seismic fragility capacity. To evaluate the seismic capacity of interface pipes in the isolated NPP, firstly, we should define the failure mode and failure criteria of critical pipe components. Hence, in this study, we performed the dynamic tests of elbow components which were installed in a seismically isolated NPP, and evaluated the ultimate failure mode and failure criteria by using the test results. To do this, we manufactured 25 critical elbow component specimens and performed cyclic loading tests under the internal pressure condition. The failure mode and failure criteria of a pipe component will be varied by the design parameters such as the internal pressure, pipe diameter, loading type, and loading amplitude. From the tests, we assessed the effects of the variation parameters onto the failure criteria. For the tests, we generated the seismic input protocol of relative displacement between the ends of elbow component. In this paper, elbow in piping system was defined as a fragile element and numerical model was updated by component test. Failure mode of piping component under seismic load was defined by the dynamic tests of ultimate pipe capacity. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. In this study, the dynamic tests were performed for the elbow components which were installed in an actual NPPs, and the ultimate failure mode and failure criteria were also evaluated by using the test results.

  17. Status of the seismic upgrading programme at Mochovce NPP

    International Nuclear Information System (INIS)

    Zajicek, T.; Dolnik, R.; Stevko, M.

    2001-01-01

    The paper provides an overview of the seismic characterisation of the Mochovce site in Slovakia. Particularly, emphasis is given to differences between the original siting and design procedures and the re-evaluation approach, much more based on the data from the micro-earthquake monitoring system installed at the site. Details are also provided for the seismic monitoring of the buildings, as confirmation of the design assumptions. (author)

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

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

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

  19. New trends in designing NPP control boards

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.

    1981-01-01

    A short analytical summary of the latest developments and future trends in designing NPP control boards is given. The designs of the Westinghause and the Hynkley-Point NPP control boards are described in detail. The essence of the advanced control board concept consists , firstly, in expanded use of computer-controlled displays for the sake of reducing the content of unimportant information presented to an operator, and, secondary, in better account of human possibilities to convert the NPP operation information into a more suitable form. An enlarged use of the direct digital reactor control utilizing microprocessors is expected. Besides, the employment of full-scale control board mock-ups and information desks as well as testing newly-developed control boards at computer reactor simulators are concluded to be used at all-growing rate [ru

  20. Seismic qualification tests of fans of the NPP of Laguna Verde U-1 and U-2

    International Nuclear Information System (INIS)

    Jarvio C, G.; Garcia H, E. E.; Arguelles F, R.; Vela H, A.; Naranjo U, J. L.

    2013-10-01

    This work presents the results of the seismic qualification tests applied to the fans that will be installed in the control panels of the three divisions of the diesel generators of the nuclear power plant (NPP) of Laguna Verde, Unit-1 and Unit-2. This seismic qualification process of the fans was carried out using two specimens that were tested in the seismic table (vibrating) of the Engineering Institute of Universidad Nacional Autonoma de Mexico (UNAM), in accordance with the requirements of the standard IEEE 344-1975, to satisfy the established requirements of seismic qualification in the technical specifications and normative documents required by the nuclear standards, in order to demonstrate its application in the diesel generators Divisions I, II and III of the NPP. The seismic qualification tests were developed on specimens that were retired of the NPP of Laguna Verde recently with a service life of 7.75 years. (Author)

  1. OVERVIEW ON BNL ASSESSMENT OF SEISMIC ANALYSIS METHODS FOR DEEPLY EMBEDDED NPP STRUCTURES

    International Nuclear Information System (INIS)

    XU, J.; COSTANTINO, C.; HOFMAYER, C.; GRAVES, H.

    2007-01-01

    A study was performed by Brookhaven National Laboratory (BNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (USNRC), to determine the applicability of established soil-structure interaction analysis methods and computer programs to deeply embedded and/or buried (DEB) nuclear power plant (NPP) structures. This paper provides an overview of the BNL study including a description and discussions of analyses performed to assess relative performance of various SSI analysis methods typically applied to NPP structures, as well as the importance of interface modeling for DEB structures. There are four main elements contained in the BNL study: (1) Review and evaluation of existing seismic design practice, (2) Assessment of simplified vs. detailed methods for SSI in-structure response spectrum analysis of DEB structures, (3) Assessment of methods for computing seismic induced earth pressures on DEB structures, and (4) Development of the criteria for benchmark problems which could be used for validating computer programs for computing seismic responses of DEB NPP structures. The BNL study concluded that the equivalent linear SSI methods, including both simplified and detailed approaches, can be extended to DEB structures and produce acceptable SSI response calculations, provided that the SSI response induced by the ground motion is very much within the linear regime or the non-linear effect is not anticipated to control the SSI response parameters. The BNL study also revealed that the response calculation is sensitive to the modeling assumptions made for the soil/structure interface and application of a particular material model for the soil

  2. Assessment of effectiveness of anti-seismic measures in stabilization project of ChNPP shelter object

    International Nuclear Information System (INIS)

    Kondrat'ev, S.N.; Kritskij, V.B.; Ryzhov, D.I.; Shugajlo, A.P.; Shugajlo, Al.P.; Prabkhakara, M.

    2004-01-01

    The major factors, which may lead to the collapse of the Shelter object (SO) civil structures, are extreme natural phenomena and among them earthquake. In order to raise the resistance of the SO civil structure to seismic and other significant loads and to reduce the risk of their collapse ChNPP requested KSK Consortium to develop the SO Detailed Design for stabilization. At the present work the results of assessment of anti-seismic measures are given based on results of a technical review of the Detailed Design

  3. Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000TM NPP

    International Nuclear Information System (INIS)

    Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San Vicente, J. L.

    2011-01-01

    Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000 T M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

  4. Sensitivity Analysis on Elbow Piping Components in Seismically Isolated NPP under Seismic Loading

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Kun; Hahm, Dae Gi; Kim, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    In this study, the FE model is verified using specimen test results and simulation with parameter variations are conducted. Effective parameters will randomly sampled and used as input values for simulations to be applied to the fragility analysis. pipelines are representative of them because they could undergo larger displacements when they are supported on both isolated and non-isolated structures simultaneously. Especially elbows are critical components of pipes under severed loading conditions such as earthquake action because strain is accumulated on them during the repeated bending of the pipe. Therefore, seismic performance of pipe elbow components should be examined thoroughly based on the fragility analysis. Fragility assessment of interface pipe should take different sources of uncertainty into account. However, selection of important sources and repeated tests with many random input values are very time consuming and expensive, so numerical analysis is commonly used. In the present study, finite element (FE) model of elbow component will be validated using the dynamic test results of elbow components. Using the verified model, sensitivity analysis will be implemented as a preliminary process of seismic fragility of piping system. Several important input parameters are selected and how the uncertainty of them are apportioned to the uncertainty of the elbow response is to be studied. Piping elbows are critical components under cyclic loading conditions as they are subjected large displacement. In a seismically isolated NPP, seismic capacity of piping system should be evaluated with caution. Seismic fragility assessment preliminarily needs parameter sensitivity analysis about the output of interest with different input parameter values.

  5. Seismic upgrading of the spent fuel storage building at Kozloduy NPP

    International Nuclear Information System (INIS)

    Alexandrov, A.; Borov, V.; Jordanov, M.; Karamanski, T.; Mihaylov, K.

    2001-01-01

    The Spent Fuel Storage Building at Kozloduy NPP site has been analysed for new review level earthquake with 0.2 g peak ground acceleration (compared to the initial design basis earthquake with 0.1 g PGA). The preliminary seismic analysis of the existing building structure using the 5% site specific response spectrum showed the need of seismic structural upgrading. Two upgrading concepts were evaluated on the basis of several factors. The main factor considered was preventing the collapse of the hall structure and the travelling cranes on the fuel storage area during and after a SSE. A three dimensional finite element model was created for the investigation of the seismic response of the existing structure and for the design of the building upgrading. The modelling of the heavy travelling crane and its sub-crane structure was one of the key points. Different configurations of the new upgrading and strengthening structures were investigated. Some interesting conclusions have been drawn from the experience in analysing and upgrading of such a complex industrial structure, comprised of elements with substantial differences in material, rigidity, construction and general behaviour. (author)

  6. Checking of seismic and tsunami hazard for coastal NPP of Chinese continent after Fukushima nuclear accident

    Institute of Scientific and Technical Information of China (English)

    Chang Xiangdong; Zhou Bengang; Zhao Lianda

    2013-01-01

    A checking on seismic and tsunami hazard for coastal nuclear power plant (NPP) of Chinese continent has been made after Japanese Fukushima nuclear accident caused by earthquake tsunami.The results of the checking are introduced briefly in this paper,including the evaluations of seismic and tsunami hazard in NPP siting period,checking results on seismic and tsunami hazard.Because Chinese coastal area belongs to the continental shelf and far from the boundary of plate collision,the tsunami hazard is not significant for coastal area of Chinese continent.However,the effect from tsunami still can' t be excluded absolutely since calculated result of Manila trench tsunami source although the tsunami wave is lower than water level from storm surge.The research about earthquake tsunami will continue in future.The tsunami warning system and emergency program of NPP will be established based on principle of defense in depth in China.

  7. Assessment of NPP safety taking into account seismic and engineering-geological factors

    International Nuclear Information System (INIS)

    Yakovlev, E.A.

    1990-01-01

    Consideration is given to the problem of probabilistic analysis of NPP safety with account of risk of destructive effect of earthquakes and the danger of accidental geological processes (diapirism, karst etc.) under NPP operation. It is shown that account of seismic and engineering-geological (engineering-seismological) risk factors in probabilistic analysis of safety enables to perform anticipatory analysis of behaviour of principle plant objects and to improve safety of their operation by revealing the most unstable elements of geotechnical system forming the main contribution to the total NPP risk

  8. Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

    International Nuclear Information System (INIS)

    Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

    2011-01-01

    Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

  9. Seismic soil-structure-equipment interaction analysis of unit 5/6, Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M [Bulgarian Academy of Sciences, Central Laboratory for Seismic Mechanics and Earthquake Engineering, Sofia (Bulgaria)

    1995-07-01

    This research project is aimed to analyse problems of soil-structure-equipment interaction under seismic excitation in case of Kozloduy NPP. Reevaluation and upgrading of Kozloduy NPP has started after 1977 Vrancea earthquake. New Safe Shutdown Earthquake (SSE) level was defined, upgrading most of structural equipment was performed, seismic instrumentation was installed. New investigations were initiated after 1990 IAEA mission visited the site. A comprehensive site confirmation project was started with a subsequent structural and equipment reevaluation and upgrading. This work deals with Units 5 and 6 of WWER-1000 type only.

  10. Seismic analysis of the safety related piping and PCLS of the WWER-440 NPP

    International Nuclear Information System (INIS)

    Berkovski, A.M.; Kostarev, V.V.; Schukin, A.J.; Boiadjiev, Z.; Kostov, M.

    2001-01-01

    This paper presents the results of seismic analysis of Safety Related Piping Systems of the typical WWER-440 NPP. The methodology of this analysis is based on WANO Terms of Reference and ASME BPVC. The different possibilities for seismic upgrading of Primary Coolant Loop System (PCLS) were considered. The first one is increasing of hydraulic snubber units and the second way is installation of limited number of High Viscous Dampers (HVD). (author)

  11. ASSESSMENT OF SEISMIC ANALYSIS METHODOLOGIES FOR DEEPLY EMBEDDED NPP STRUCTURES

    International Nuclear Information System (INIS)

    XU, J.; MILLER, C.; COSTANTINO, C.; HOFMAYER, C.; GRAVES, H. NRC.

    2005-01-01

    Several of the new generation nuclear power plant designs have structural configurations which are proposed to be deeply embedded. Since current seismic analysis methodologies have been applied to shallow embedded structures (e.g., ASCE 4 suggest that simple formulations may be used to model embedment effect when the depth of embedment is less than 30% of its foundation radius), the US Nuclear Regulatory Commission is sponsoring a program at the Brookhaven National Laboratory with the objective of investigating the extent to which procedures acceptable for shallow embedment depths are adequate for larger embedment depths. This paper presents the results of a study comparing the response spectra obtained from two of the more popular analysis methods for structural configurations varying from shallow embedment to complete embedment. A typical safety related structure embedded in a soil profile representative of a typical nuclear power plant site was utilized in the study and the depths of burial (DOB) considered range from 25-100% the height of the structure. Included in the paper are: (1) the description of a simplified analysis and a detailed approach for the SSI analyses of a structure with various DOB, (2) the comparison of the analysis results for the different DOBs between the two methods, and (3) the performance assessment of the analysis methodologies for SSI analyses of deeply embedded structures. The resulting assessment from this study has indicated that simplified methods may be capable of capturing the seismic response for much deeper embedded structures than would be normally allowed by the standard practice

  12. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  13. Seismic response analyses of turbine hall and electrical building of RBMK-1000 MW type NPP

    International Nuclear Information System (INIS)

    Jordanov, M.J.; Karparov, K.T.

    2003-01-01

    This paper addresses results obtained during the study of turbine hall and electrical building of RBMK-1000 MW pair units at Leningradskaya NPP (LNPP) for seismic event. The study was performed in the frame of the Coordinated Research Program of the International Atomic Agency (IAEA) on Safety of RBMK type Nuclear Power Plants (NPP) in Relation of External Events. A 3-D finite element model of Main Building Complex was developed and seismic response analyses were performed taking into account the soil-structure interaction (SSI). The standard mode superposition method was used for evaluation of dynamic response of structure in time domain. The structure was assumed surface founded at the basemat level. Seismic response analyses were carried out considering shear wave propagation pattern for the input motion. The in-structure time histories and response spectra were generated in referenced locations. Conclusions are drawn for the reliability of the structural response evaluation considering the soil-structure interaction effects. (author)

  14. Civil Works Seismic Designs

    International Nuclear Information System (INIS)

    1985-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. This rule defines: - the parameters characterizing the design seismic motions - the calculation methods - the mathematical schematization principles on which calculations are based - the use of the seismic response for the structure checking - the content of the documents to be presented

  15. Seismic response and fragility evaluation for an Eastern US NPP including soil-structure interaction effects

    International Nuclear Information System (INIS)

    Ghiocel, Dan M.; Wilson, Paul R.; Thomas, Gary G.; Stevenson, John D.

    1998-01-01

    The paper discusses methodological aspects involved in a probabilistic seismic soil-structure interaction (SSI) analysis for a Seismic Probabilistic Risk Assessment (SPRA) review. An example of an Eastern US nuclear power plant (NPP) is presented. The approach presented herein follows the current practice of the Individual Plant Examination for External Events (IPEEE) program in the US. The NPP is founded on a relatively soft soil deposit, and thus the SSI effects on seismic responses are significant. Probabilistic models used for the idealization of the seismic excitation and the surrounding soil deposit are described. Using a lognormal format, computed random variability effects were combined with those proposed in the SPRA methodology guidelines. Probabilistic floor response spectra and structural fragilities for different NPP buildings were computed. Structural capacities were determined following the current practice which assumes independent median safety factors for strength and inelastic absorption. Limitations of the IPEEE practice for performing SPRA are discussed and alternate procedures, more rigorous and simple to implement, are suggested

  16. Seismic response analysis of a piping system subjected to multiple support excitations in a base isolated NPP building

    International Nuclear Information System (INIS)

    Surh, Han-Bum; Ryu, Tae-Young; Park, Jin-Sung; Ahn, Eun-Woo; Choi, Chul-Sun; Koo, Ja Choon; Choi, Jae-Boong; Kim, Moon Ki

    2015-01-01

    Highlights: • Piping system in the APR 1400 NPP with a base isolation design is studied. • Seismic response of piping system in base isolated building are investigated. • Stress classification method is examined for piping subjected to seismic loading. • Primary stress of piping is reduced due to base isolation design. • Substantial secondary stress is observed in the main steam piping. - Abstract: In this study, the stress response of the piping system in the advanced power reactor 1400 (APR 1400) with a base isolation design subjected to seismic loading is addressed. The piping system located between the auxiliary building with base isolation and the turbine building with a fixed base is considered since it can be subjected to substantial relative support movement during seismic events. First, the support responses with respect to the base characteristic are investigated to perform seismic analysis for multiple support excitations. Finite element analyses are performed to predict the piping stress response through various analysis methods such as the response spectrum, seismic support movement and time history method. To separately evaluate the inertial effect and support movement effect on the piping stress, the stress is decomposed into a primary and secondary stress using the proposed method. Finally, influences of the base isolation design on the piping system in the APR 1400 are addressed. The primary stress based on the inertial loading is effectively reduced in a base isolation design, whereas a considerable amount of secondary stress is generated in the piping system connecting a base isolated building with a fixed base building. It is also confirmed that both the response spectrum analysis and seismic support movement analysis provide more conservative estimations of the piping stress compared to the time history analysis

  17. Seismic strengthening of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Stoyanov, G.; Jordanov, M.

    2005-01-01

    This paper presents results obtained during the upgrading design of overhead roads (OHR) between WWER-1000 MW Reactor Units at Kozloduy NPP. In order to avoid the deficiencies of OHR seismic capacity different approaches were developed based on the site and structure specifics. Overhead roads are precasted RC structures. They consist of pedestrian gallery and pipeline RC box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated at their as-is status and a seismic upgrading of the structure is designed. The analysis of the upgraded structure is performed for Review Level Earthquake (RLE). Soil-Structure Interaction (SSI) effects are taken into account through equivalent soil springs with frequency adjusted stiffnesses. The upgraded structure is checked for conformance with the specially developed technical design specification based on International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The existing site conditions and Owner's requirements are taken into account during development of the upgrading. The proposed upgrading measures can be divided in two major categories global and local. Special attention is paid to improvement of the ductile behavior of the structure through new detailing and upgrading of existing connection. These measures are grouped in two final design concepts and after a comparative study one of them is chosen for implementation. For the upgraded structure response spectra are derived at locations where equipment is attached. (authors)

  18. Design of NPP of new generation being constructed at the Novovoronezh NPP site

    International Nuclear Information System (INIS)

    Afrov, A.; Berkovich, V.; Generalov, V.; Dragunov, Yu.; Krushelnitsky, V.

    1999-01-01

    The design of a new generation NPP is described, underscoring advances in physical attributes and passive safety systems based on experiences with earlier designs at operating NPPs. This paper elaborates on systems for handling and storing radioactive wastes, on refinements in containment measures and on experimental and analytic validation of critical design factors. (author)

  19. Risk based seismic design criteria

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1999-01-01

    In order to develop a risk based seismic design criteria the following four issues must be addressed: (1) What target annual probability of seismic induced unacceptable performance is acceptable? (2) What minimum seismic margin is acceptable? (3) Given the decisions made under Issues 1 and 2, at what annual frequency of exceedance should the safe-shutdown-earthquake (SSE) ground motion be defined? (4) What seismic design criteria should be established to reasonably achieve the seismic margin defined under Issue 2? The first issue is purely a policy decision and is not addressed in this paper. Each of the other three issues are addressed. Issues 2 and 3 are integrally tied together so that a very large number of possible combinations of responses to these two issues can be used to achieve the target goal defined under Issue 1. Section 2 lays out a combined approach to these two issues and presents three potentially attractive combined resolutions of these two issues which reasonably achieves the target goal. The remainder of the paper discusses an approach which can be used to develop seismic design criteria aimed at achieving the desired seismic margin defined in resolution of Issue 2. Suggestions for revising existing seismic design criteria to more consistently achieve the desired seismic margin are presented. (orig.)

  20. Soil-structural interaction analysis of RBMK type NPP for seismic event. Progress report. From 1 July 1998 - 30 June 1999

    International Nuclear Information System (INIS)

    1999-01-01

    The objective of the project is to assess the structural behavior and safety capacity of a RBMK-1000 MW Main Building Complex under critical combination of loads including seismic events. This project is part of the Coordinated Research Program carried out by International Atomic Energy Agency on safety of RBMK Type Nuclear Power Plants (NPP) in Relation to External Events. The nuclear power plant considered for this study is the Sosnovy Bor NPP, located near St.Petersburg, Russia. The Soviet standard design RBMK-1000 MW type units installed in Sosnovy Bor NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1 g. The relevant response spectra are not available for reference and assessment. The new international requirements for nuclear power plants in operation require site specific seismic hazard studies as a basis for the definition of a Review Level Earthquake (RLE) for reassessment of the structures and safety related equipment Ell - As the RLE site specific seismic data is still not available, the RLE earthquake spectra for Kozloduy NPP scaled to PGA=0.1 g were used in this study. This value is intentionally chosen for comparison purposes. The Russian design requirements (if design floor response spectra are available) will be compared with the international regulations. The scope of the study is to perform a Soil-Structure Interaction (SSI) seismic response analysis of the referenced RBMK-11000 MW. Main Building Complex to evaluate the effect on the structural response of a greater than design earthquake. The analysis is focused on a realistic assessment of the structural response to a potentially higher earthquake level instead of a conservative design type analysis. Special attention is paid on the seismic response of the sub-structures in the safe shutdown path, as well as on the locations of the heavy equipment

  1. Seismic assessment of safety-related structures: laboratory testing of the pressure relief duct frame at pickering NPP

    International Nuclear Information System (INIS)

    Ghobarah, A.; Biddah, A.; Pilette, C.

    1995-01-01

    The pressure relief duct (PRD) is a Special safety System in the CANDU-PHW multi-unit nuclear power plants (NPP). It is designed to contain and direct the outflow from the reactor building to the pressure suppression and containing systems in the vacuum building. The PRD is a large elevated reinforced concrete box structure of internal width of 6.1 m, height of 7.6 m, and wall thickness of 0.6 m. The PRD is 662 m long and is supported every 22 m by concrete frames of height of 21 m. Typical frame members are 1.8 m in depth and width. A representative elevation of the frame is presented. The section of the PRD under investigation was designed and constructed before the current seismic design codes were in effect. An assessment of the PRD structure when subjected to various levels of ground motion has shown that the frame has a limited seismic withstand capacity. Its seismic performance is dependent on the ductility of the beams and on the ability of the beam-column joint to transfer bending moments and shear forces. The objectives of this study are to provide the data to validate the frame analysis results through laboratory testing of a scaled specimen of the beam-column joint, and to compare the observed response with the response of a beam-column joint when the shear reinforcement is detailed according to current seismic design codes. (author). 3 refs., 10 figs

  2. Review of studies pertaining to the seismic input at Paks NPP

    International Nuclear Information System (INIS)

    Muzzi, F.

    1995-01-01

    This report refers to the examination performed on the available material relevant for the seismic input estimate for the Paks NPP, within the frame of the IAEA benchmark study for the seismic analysis and testing of the existing NPPs. The aim of the report is to provide an expert judgement about the quantity and quality of the data and studies performed. The first chapter describes the sources of the data set examined, the second involves the criteria followed in the judgment. The third chapter contains the detailed opinion on the content of the data set, the conclusion and suggestions are reported in chapter four

  3. Optimal organization of structural analysis and site inspection for the seismic requalification of Paks NPP

    International Nuclear Information System (INIS)

    Contri, P.

    1996-01-01

    The analysis described in this report deals with a numerical procedure aimed for the assessment of a methodology for the optimal organization of data collection, in the context of seismic requalification of structures and components of existing nuclear power plants. The presented procedure has quite a general application and an example was chosen for the Paks NPP where seismic requalification is in progress. The assessment was carried out in reference to the following main tasks: structure and soil data analysis; numerical model generation; deterministic dynamic analysis description; reliability analysis framework discussion; transfer function calculation via response surface approach; and the sensitivity evaluation

  4. Position paper: Seismic design criteria

    International Nuclear Information System (INIS)

    Farnworth, S.K.

    1995-01-01

    The purpose of this paper is to document the seismic design criteria to be used on the Title 11 design of the underground double-shell waste storage tanks and appurtenant facilities of the Multi-Function Waste Tank Facility (MWTF) project, and to provide the history and methodologies for determining the recommended Design Basis Earthquake (DBE) Peak Ground Acceleration (PGA) anchors for site-specific seismic response spectra curves. Response spectra curves for use in design are provided in Appendix A

  5. A SEISMIC DESIGN OF NUCLEAR REACTOR BUILDING STRUCTURES APPLYING SEISMIC ISOLATION SYSTEM IN A HIGH SEISMICITY REGION –A FEASIBILITY CASE STUDY IN JAPAN-

    Directory of Open Access Journals (Sweden)

    TETSUO KUBO

    2014-10-01

    Full Text Available A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1 the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2 the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3 the responses of isolated reactor building fall below the range of the prescribed criteria.

  6. A seismic design of nuclear reactor building structures applying seismic isolation system in a seismicity region-a feasibility case study in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Tetsuo [The University of Tokyo, Tokyo (Japan); Yamamoto, Tomofumi; Sato, Kunihiko [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Jimbo, Masakazu [Toshiba Corporation, Yokohama (Japan); Imaoka, Tetsuo [Hitachi-GE Nuclear Energy, Ltd., Hitachi (Japan); Umeki, Yoshito [Chubu Electric Power Co. Inc., Nagoya (Japan)

    2014-10-15

    A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB) is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1) the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2) the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3) the responses of isolated reactor building fall below the range of the prescribed criteria.

  7. Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000{sup T}M NPP

    Energy Technology Data Exchange (ETDEWEB)

    Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San vicente, J. L.

    2011-07-01

    Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000{sup T}M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

  8. Structure study and design of Qinshan NPP PCCV

    International Nuclear Information System (INIS)

    Xia Zufeng; Xu Yongzhi; Wang Tianzhen; Wu Jibiao

    1993-02-01

    The design process of Qinshan NPP (nuclear power plant) PCCV (prestressed concrete containment vessel) is summarized. The tendon test, structural description, design bases and analysis method are introduced. The arrangement for preventing concrete from cracking and design features of post-tensioning system and steel liner are presented. The results of model test and non-linear analysis for ultimate load in Qinshan NPP PCCV are also given. Through the integrity test of PCCV, it shows that the test values are in agreement with predicted values, the structure is excellent and the performance of leak tightness conforms to the safety requirements

  9. Overview of seismic resistant design of Indian Nuclear Power Plants

    International Nuclear Information System (INIS)

    Sharma, G.K.; Hawaldar, R.V.K.P.; Vinod Kumar

    2007-01-01

    Safe operation of a Nuclear Power Plant (NPP) is of utmost importance. NPPs consist of various Structure, System and Equipment (SS and E) that are designed to resist the forces generated due to a natural phenomenon like earthquake. An earthquake causes severe oscillatory ground motion of short duration. Seismic resistant design of SS and E calls for evaluation of effect of severe ground shaking for assuring the structural integrity and operability during and after the occurrence of earthquake event. Overall exercise is a multi-disciplinary approach. First of standardized 220 MWe design reactor is Narora Atomic Power Station. Seismic design was carried out as per state of art then, for the first time. The twelve 220 MWe reactors and two 540 MWe reactors designed since 1975 have been seismically qualified for the earthquake loads expected in the region. Seismic design of 700 MWe reactor is under advanced stage of finalization. Seismic re-evaluation of six numbers of old plants has been completed as per latest state of art. Over the years, expertise have been developed at Nuclear Power Corporation of India Limited, Bhabha Atomic Research Centre, prominent educational institutes, research laboratories and engineering consultants in the country in the area of seismic design, analysis and shake table testing. (author)

  10. Seismic design considerations for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

    2001-01-01

    During the last few decades, there have been considerable advances in the field of a seismic design of nuclear structures and components housed inside a Nuclear power Plant (NPP). The seismic design and qualification of theses systems and components are carried out through the use of well proven and established theoretical as well as experimental means. Many of the related research works pertaining to these methods are available in the published literature, codes, guides etc. Contrary to this, there is very little information available with regards to the seismic design aspects of the nuclear fuel cycle facilities. This is probably on account of the little importance attached to these facilities from the point of view of seismic loading. In reality, some of these facilities handle a large inventory of radioactive materials and, therefore, these facilities must survive during a seismic event without giving rise to any sort of undue radiological risk to the plant personnel and the public at large. Presented herein in this paper are the seismic design considerations which are adopted for the design of nuclear fuel cycle facilities in India. (author)

  11. A Study on the Development of Prototype Seismic Isolation Device for NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hongpyo; Cho, Myungsug; Kim, Sunyong; Lee, Yonghee; Kang Kyunghun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-05-15

    Korean nuclear power plants have been and still are based on seismic resistance design including all of the natural disasters. However, in regions of high seismic hazard, seismic isolation technology is needed to guarantee the seismic safety on nuclear power plants. To achieve this purpose, the research and development of seismic isolation system for the construction in high seismicity area is on-going in Korea. In this study, prototype seismic isolation devices as mentioned above are developed and tested to identify the basic shear and compressive characteristics of them. In this study, assessment performance of basic characteristics on the prototype LRB and EQS seismic isolation for nuclear power plant structures is employed to compare with design values. Based on the test results of compression and shear characteristics, it is judged that they meet the measuring efficiency range conditions which are presented in ISO 22762 and AASHOT guide specification. Therefore, prototype seismic isolation devices like LRB and EQS developed in this study can be expected to be used as reference data when designing a seismic isolation system for nuclear power plant structures in the future.

  12. Development of seismic safety reevaluation procedure considering the ageing of NPP facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Kue [Jeonju Univ., Cheonju (Korea, Republic of); Kim, J. M. [Cheonnam National Univ., Gwangju (Korea, Republic of); Kim, Y. S.; Cheong, S. H.; Kim, I. S.; Lee, M. G.; Kim, D. O. [Andong National Univ., Andong (Korea, Republic of); Lee, G. H. [Mokpo National Maritime Univ., Mokpo (Korea, Republic of)

    2003-03-15

    There are three of Nuclear Power Plants subject to the USI A-46 in Korea, including Kori No 1 and No 2 and Wolsung No 1. For the sake of resolution of the issue the possibility of adopting the GIP developed by the SQUG in USA is very high. In relation to the issue, this study addresses some technical improvements of the GIP including sloshing analysis based on multiple modes, seismic retrofit of cabinet for reduction of ICRS and modification of IRS depending on damping ratio. Dominant degradation factor and its affects NPP concrete elements are reviewed : chloride induced corrosion, carbonation of concrete elements, freezing and thawing of concrete elements, chemical and biological process, crack affect on concrete degradation. Various technical reports and papers about age-related degradation are reviewed for identification of degradation properties of NPP structures and components and degradation trend in NPP structures and components. This report summarizes numerical model for concrete degradation and development procedure of numerical models for concrete degradation. This report proposes the research necessity for performance evaluation of degraded concrete structure and selection of element for further study.

  13. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3E. Kozloduy NPP units 5/6: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to floor response spectra of Kozloduy NPP; calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating WWER-type NPPs; analysis of design floor response spectra and testing of the electrical systems; experimental investigations and seismic analysis Kozloduy NPP; testing of components on the shaking table facilities and contribution to full scale dynamic testing of Kozloduy NPP; seismic evaluation of the main steam line, piping systems, containment pre-stressing and steel ventilation chimney of Kozloduy NPP

  14. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3H. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  15. Core design methodology and software for Temelin NPP

    International Nuclear Information System (INIS)

    Havluj, F; Hejzlar, J.; Klouzal, J.; Stary, V.; Vocka, R.

    2011-01-01

    In the frame of the process of fuel vendor change at Temelin NPP in the Czech Republic, where, starting since 2010, TVEL TVSA-T fuel is loaded instead of Westinghouse VVANTAGE-6 fuel, new methodologies for core design and core reload safety evaluation have been developed. These documents are based on the methodologies delivered by TVEL within the fuel contract, and they were further adapted according to Temelin NPP operational needs and according to the current practice at NPP. Along with the methodology development the 3D core analysis code ANDREA, licensed for core reload safety evaluation in 2010, have been upgraded in order to optimize the safety evaluation process. New sequences of calculations were implemented in order to simplify the evaluation of different limiting parameters and output visualization tools were developed to make the verification process user friendly. Interfaces to the fuel performance code TRANSURANUS and sub-channel analysis code SUBCAL were developed as well. (authors)

  16. Study on integrated design and analysis platform of NPP

    International Nuclear Information System (INIS)

    Lu Dongsen; Gao Zuying; Zhou Zhiwei

    2001-01-01

    Many calculation software have been developed to nuclear system's design and safety analysis, such as structure design software, fuel design and manage software, thermal hydraulic analysis software, severe accident simulation software, etc. This study integrates those software to a platform, develops visual modeling tool for Retran, NGFM90. And in this platform, a distribution calculation method is also provided for couple calculation between different software. The study will improve the design and analysis of NPP

  17. A basis for standardized seismic design (SSD) for nuclear power plants/critical facilities

    International Nuclear Information System (INIS)

    O'Hara, T.F.; Jacobson, J.P.; Bellini, F.X.

    1991-01-01

    US Nuclear Power Plants (NPP's) are designed, engineered and constructed to stringent standards. Their seismic adequacy is assured by compliance with regulatory standards and demonstrated by both probabilistic risk assessments (PRAs) and seismic margin studies. However, present seismic siting criteria requires improvement. Proposed changes to siting criteria discussed here will provide a predictable licensing process and a stable regulatory environment. Two recent state-of-the-art studies evaluate the seismic design for all eastern US (EUS) NPP'S: a Lawrence Livermore National Labs study (LLNL, 1989) funded by the NRC and similar research by the Electric Power Research Institute (EPRI, 1989) supported by the utilities. Both confirm that Appendix A 10CFR Part 100 has not provided consistent seismic design levels for all sites. Standardized Seismic Design (SSD) uses a probabilistic framework to accommodate alternative deterministic interpretations. It uses seismic hazard input from EPRI or LLNL to produce consistent bases for future seismic design. SSD combines deterministic and probabilistic insights to provide a comprehensive approach for determining a future site's acceptable seismic design basis

  18. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

    This paper deals with the method used in France for the PWR nuclear plants to derive locations and types of supports of auxiliary and secondary piping systems taking earthquake in account. The successive steps of design are described, then the seismic computation method and its particular conditions of applications for piping are presented. The different types of support (and especially seismic ones) are described and also their conditions of installation. The method used to compare functional tests results and computation results in order to control models is mentioned. Some experiments realised on site or in laboratory, in order to validate models and methods, are presented [fr

  19. Displacement Based Seismic Design Criteria

    International Nuclear Information System (INIS)

    Costello, J.F.; Hofmayer, C.; Park, Y.J.

    1999-01-01

    The USNRC has initiated a project to determine if any of the likely revisions to traditional earthquake engineering practice are relevant to seismic design of the specialized structures, systems and components of nuclear power plants and of such significance to suggest that a change in design practice might be warranted. As part of the initial phase of this study, a literature survey was conducted on the recent changes in seismic design codes/standards, on-going activities of code-writing organizations/communities, and published documents on displacement-based design methods. This paper provides a summary of recent changes in building codes and on-going activities for future codes. It also discusses some technical issues for further consideration

  20. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4F. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  1. Design and operating condition. Consequences for EDF NPP's chemistry

    International Nuclear Information System (INIS)

    Bretelle, Jean-Luc; Stutzmann, Agnes; Nordmann, Francis

    2009-01-01

    Since the beginning of the French nuclear programme in 1977, four major types of design have been commissioned, the fifth one being under construction (EPR). The paper points out advantages and drawbacks of chemistry choices for the primary, secondary and tertiary systems in French NPP, for each design particularity and it describes the corresponding operating conditions. Chemistry option proposals are drawn for the future of the French fleet, taking into account the material behaviour and the operation improvement. (orig.)

  2. Establishing seismic design criteria to achieve an acceptable seismic margin

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1997-01-01

    In order to develop a risk based seismic design criteria the following four issues must be addressed: (1) What target annual probability of seismic induced unacceptable performance is acceptable? (2). What minimum seismic margin is acceptable? (3) Given the decisions made under Issues 1 and 2, at what annual frequency of exceedance should the Safe Shutdown Earthquake ground motion be defined? (4) What seismic design criteria should be established to reasonably achieve the seismic margin defined under Issue 2? The first issue is purely a policy decision and is not addressed in this paper. Each of the other three issues are addressed. Issues 2 and 3 are integrally tied together so that a very large number of possible combinations of responses to these two issues can be used to achieve the target goal defined under Issue 1. Section 2 lays out a combined approach to these two issues and presents three potentially attractive combined resolutions of these two issues which reasonably achieves the target goal. The remainder of the paper discusses an approach which can be used to develop seismic design criteria aimed at achieving the desired seismic margin defined in resolution of Issue 2. Suggestions for revising existing seismic design criteria to more consistently achieve the desired seismic margin are presented

  3. Seismic analysis of design

    International Nuclear Information System (INIS)

    Jehlicka, P.

    1980-01-01

    The determination of the dynamic response of nuclear power plants is a necessary part of safe design against earthquake, or against other additional vibrational loading. The determination of these dynamic loads caused by external excitation is a requirement in calculating the related material loading on the structures. The purpose of this lecture is to present a general survey of analytical methods to determine the response of structural and mechanical equipment to earthquake. The main problems which complicate structural-dynamic calculations will be discussed. The necessity to control input parameters and the possibility to calculate with simplified methods will be pointed out. (orig./RW)

  4. Seismic design and qualification for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This safety guide, which supplements the IAEA Code on the Safety of Nuclear Power Plants (NPP); Design (IAEA Safety Series No.50-C-D (Rev.1)), forms part of the Agency's programme, referred to as the NUSS programme, for establishing Codes and Guides relating to land based stationary thermal neutron power plants. The present Guide was originally issued in 1979 as Safety Guide 50-SG-S2 within the series of NUSS guides for the siting of NPP, extending seismic considerations from Safety Guide 50-SG-S1 into the design and verification field. During the revision phase in 1988-1990, this emphasis on design aspects was confirmed and consequently the Guides have been reclassified as a design Guide with the corresponding identification number 50-SG-D15. The general character of the Guide has not been changed an it still relates strongly to 50-SG-S1, which gives guidance on how to determine design basis ground motion for a NPP at a given site

  5. Seismic analysis of a NPP reactor building using spectrum-compatible power spectral density functions

    International Nuclear Information System (INIS)

    Venancio Filho, F.; DeCarvalho Santos, S.H.; Joia, L.A.

    1987-01-01

    A numerical methodology to obtain Power Spectral Density Functions (PSDF) of ground accelerations, compatible with a given design response spectrum is presented. The PSDF's are derived from the statistical analysis of the amplitudes of the frequency components in a set of artificially generated time-histories matching the given spectrum. A so obtained PSDF is then used in the stochastic analysis of a NPP Reactor Building. The main results of this analysis are compared with the ones obtained by deterministic methods

  6. Belene NPP

    International Nuclear Information System (INIS)

    Tsvetanov, P.

    1990-01-01

    The book presents the main results of the studies of the Bulgarian Academy of Sciences (BAS) on the construction of a new nuclear power plant at Belene on the Danube river. The programme of the studies comprises five areas: the socio-economic and energy development and the necessity of the commissioning; a technical project and design level of the equipment (safety, radioactivity control, waste disposal and economic efficiency of the power plant); the seismic properties of the construction site; the corresponding risk and design features of the plant; the ecological impacts of the NPP and public opinion. The studies in the different areas have been carried out by independent teams, fully responsible for the formulated topical conclusions. The general opinion of the BAS voiced in the book is that the construction of Belene NPP is not sufficiently substantiated and is considered unacceptable. 94 refs., 53 fig., 56 tabs. (R.Ts.)

  7. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  8. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3G. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  9. NPP Engineering and Servicing / Design Analysis Department

    International Nuclear Information System (INIS)

    Sik, J.

    2006-01-01

    The article provides an overview of the activities of the SKODA JS's Design Analysis Department performed recently in the fields of reactor physics, shielding physics, thermal hydraulics and mechanical structure stresses and life analysis. (orig.)

  10. Probabilistic site dependent design spectra for a NPP

    International Nuclear Information System (INIS)

    Chavez, M.; Arroyo, M.; Romo, M.P.

    1985-01-01

    A methodology is proposed to compute the design spectra for a NPP site. Near field earthquakes are included by using an appropriately scaled sample of response spectra. Site effects are considered through a probabilistic site response analysis in the frequency domain which considers nonlinear behaviour of soils. The uncertainties of the soil shear modulus, G, are introduced by using Rosenblueth's point estimates. Strong motion duration is treated by using sensitivity analysis. The procedure is applied to a NPP site and the results are: a) the USNCR R.G.1.60 underestimate the spectral amplitudes for frequencies of interest; b) the omission of the uncertainties on the G leads to under or over-estimate the spectral amplitudes at certain frequency bands; c) the effect of considering the actual strong motion duration instead of an average value is to reduce the peak spectral amplitudes by a ten per cent. (orig.)

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3I. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  12. Seismic design features of the ACR Nuclear Power Plant

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Aziz, T.

    2003-01-01

    Through their worldwide operating records, CANDU Nuclear Power Plants (NPPs) have repeatedly demonstrated safe, reliable and competitive performance. Currently, there are fourteen CANDU 6 single unit reactors operating or under construction worldwide. Atomic Energy of Canada Limited's (AECL) Advanced CANDU Reactor - the ACR. - is the genesis of a new generation of technologically advanced reactors founded on the CANDU reactor concept. The ACR is the next step in the evolution of the CANDU product line. The ACR products (ACR-700 and ACR-1000) are based on CANDU 6 (700 MWe class) and CANDU 9 (900 MWe class) reactors, therefore continuing AECL's successful approach of offering CANDU plants that appeal to a broad segment of the power generation market. The ACR products are based on the proven CANDU technology and incorporate advanced design technologies. The ACR NPP seismic design complies with Canadian standards that were specifically developed for nuclear seismic design and also with relevant International Atomic Energy Agency (IAEA) Safety Design Standards and Guides. However, since the ACR is also being offered to several markets with many potential sites and different regulatory environments, there is a need to develop a comprehensive approach for the seismic design input parameters. These input parameters are used in the design of the standard ACR product that is suitable for many sites while also maintaining its economic competitiveness. For this purpose, the ACR standard plant is conservatively qualified for a Design Basis Earthquake (DBE) with a peak horizontal ground acceleration of 0.3g for a wide range of soil/rock foundation conditions and Ground Response Spectra (GRS). These input parameters also address some of the current technical issues such as high frequency content and near field effects. In this paper, the ACR seismic design philosophy and seismic design approach for meeting the safety design requirements are reviewed. Also the seismic design

  13. Near Regional and Site Investigations of the Temelin NPP Site

    International Nuclear Information System (INIS)

    Prachar, Ivan; Vacek, Jiri; Heralecky, Pavel

    2011-01-01

    The Temelin NPP is worldwide through heated discussion with nuclear energetic opposition. In addition this discussion goes beyond a border of the Czech Republic. On the other side, results of several international supervisions shown that Temelin NPP is fully comparable with the safest nuclear power plants in the world regarding its technical design and safety functions. This presentation deals with the near regional and site investigations of the Temelin NPP Site. It must be noted that although the Temelin site is situated in the area with low seismicity, item of seismicity is a basic argument against Temelin NPP and therefore a detail seismic hazard assessment was performed

  14. Integrated structural design of nuclear power plants for high seismic areas

    International Nuclear Information System (INIS)

    Rieck, P.J.

    1979-01-01

    A design approach which structurally interconnects NPP buildings to be located in high seismic areas is described. The design evolution of a typical 600 MWe steel cylindrical containment PWR is described as the plant is structurally upgraded for higher seismic requirements, while maintaining the original plant layout. The plant design is presented as having separate reactor building and auxiliary structures for a low seismic area (0.20 g) and is structurally combined at the foundation for location in a higher seismic area (0.30 g). The evolution is completed by a fully integrated design which structurally connects the reactor building and auxiliary structures at superstructure elevations as well as foundation levels for location in very severe seismic risk areas (0.50 g). (orig.)

  15. Seismic behavior of NPP structures subjected to realistic 3D, inclined seismic motions, in variable layered soil/rock, on surface or embedded foundations

    International Nuclear Information System (INIS)

    Jeremić, B.; Tafazzoli, N.; Ancheta, T.; Orbović, N.; Blahoianu, A.

    2013-01-01

    Highlights: • Full 3D, inclined, incoherent seismic motions used for modeling SSI of an NPP. • Analyzed effects of variable and uniform soil/rock layering profiles on SSI. • Surface and embedded foundations were modeled and differences analyzed. - Abstract: Presented here is an investigation of the seismic response of a massive NPP structures due to full 3D, inclined, un-correlated input motions for different soil and rock profiles. Of particular interest are the effects of soil and rock layering on the response and the changes of input motions (frequency characteristics) due to such layering. In addition to rock/soil layering effects, investigated are also effects of foundation embedment on dynamic response. Significant differences were observed in dynamic response of containment and internal structure founded on surface and on embedded foundations. These differences were observed for both rock and soil profiles. Select results are used to present most interesting findings

  16. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4G. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  17. Seismic design of reactors in NUCEF

    Energy Technology Data Exchange (ETDEWEB)

    Kurosaki, Akira [Mitsui Shipbuilding and Engineering Co. Ltd., Tokyo (Japan); Kuchiya, Masao; Yasuda, Naomitsu; Kitanaka, Tsutomu; Ogawa, Kazuhiko; Sakuraba, Koichi; Izawa, Naoki; Takeshita, Isao

    1997-03-01

    Basic concept and calculation method for the seismic design of the main equipment of the reactors in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) are described with actual calculation examples. The present paper is published to help the seismic design of the equipment and application of the authorization for the design and constructing of facilities. (author)

  18. Research on high level radioactive waste repository seismic design criteria

    International Nuclear Information System (INIS)

    Jing Xu

    2012-01-01

    Review seismic hazard analysis principle and method in site suitable assessment process of Yucca Mountain Project, and seismic design criteria and seismic design basis in primary design process. Demonstrated spatial character of seismic hazard by calculated regional seismic hazard map. Contrasted different level seismic design basis to show their differences and relation. Discussed seismic design criteria for preclosure phrase of high level waste repository and preference goal under beyond design basis ground motion. (author)

  19. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  20. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3F. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  1. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4E. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  2. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

    Slagis, G.C.; Moore, S.E.

    1994-01-01

    Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs

  3. Evaluation of seismic resistance of low voltage switchgear, NPP V1 Jaslovske Bohunice, Slovakia

    International Nuclear Information System (INIS)

    Zeman, P.

    1999-01-01

    During this year, company Stevenson and Associates took part in the project of evaluation of seismic resistance of NPP V-1 Jaslovske Bohunice in Slovakia. It was responsible for a part of electrical equipment, mainly for the evaluation of low voltage switchgears. There were four steps of the evaluation: Detailed Walkdown; Application of GIP-WWER Methodology; Developing, of In Cabinet Response Spectra; and Evaluation of Acceptance of Formerly Performed Relay Tests According to the Russian Standard OEG l-330.00-3). Tests performed according to the Russian Standard OAG are acceptable only if the tested subject shows just one dominant natural frequency in the significant energy frequency range. If there is no knowledge of modal properties of the tested subject (that is a frequent situation because test reports usually contain only generalized Fourier loading spectrum) the enveloping of In Cabinet Response Spectra (ICRS) in all significant energy frequency ranges by Response Spectra (RS) of harmonic signal on one arbitrary frequency. This criteria is usually not satisfied because the shake tables used for the tests are not able to produce the sufficient level of excitation in the low frequency range. It may lead to the demand for test repeating

  4. An engineering design of reactor with NPP spent fuels

    International Nuclear Information System (INIS)

    Yuan Luzheng; Shen Feng; Yang Changjiang; Dai Changnian; Jin Huajin; Li Yulun

    2005-01-01

    Study has proven that it is of practical significance to design a reactor in suitable low parameters using the spent fuels of nuclear power plant. This kind of reactor will supply, safely and economically, a clean energy for desalination of sea- water and heating supply for city residents. Based on listing main problems required to be solved when designing a reactor in suitable low parameters directly using NPP spent fuels, a preliminary design scheme with engineering feasibility is given. Some significant efforts and attempts have been made for this scheme on its core structure and main processing systems design, adopting inherent safety characteristics to the full, making the reactor as a 'foolish type' one with easy operation, safe and reliable merit to the best. (authors)

  5. Programmes design for Bohunice NPP personnel other than control room operators

    International Nuclear Information System (INIS)

    Kalincik, L.

    2002-01-01

    This paper deals with project development of training programmes for non-licenced NPP personnel-masters, field operators, maintenance and technical supporting personnel. The programme development focuses on the part stage and on the job training at NPP. Bohunice NPP belongs to plants with higher specific number of personnel per installed power capacity. This factor also influenced the choice of programmes design. Undermentioned procedure is one of various approaches to SAT exploitation for training programmes design. (author)

  6. Structural concepts and details for seismic design

    International Nuclear Information System (INIS)

    Johnson, M.W.; Smietana, E.A.; Murray, R.C.

    1991-01-01

    As a part of the DOE Natural Phenomena Hazards Program, a new manual has been developed, entitled UCRL-CR-106554, open-quotes Structural Concepts and Details for Seismic Design.close quotes This manual describes and illustrates good practice for seismic-resistant design

  7. Seismic design practices for power systems

    International Nuclear Information System (INIS)

    Schiff, A.J.

    1991-01-01

    In this paper, the evolution of seismic design practices in electric power systems is reviewed. In California the evolution had led to many installation practices that are directed at improving the seismic ruggedness of power system facilities, particularly high voltage substation equipment. The primary means for substantiating the seismic ruggedness of important, hard to analyze substation equipment is through vibration testing. Current activities include system evaluations, development of emergency response plans and their exercise, and review elements that impact the entire system, such as energy control centers and communication systems. From a national perspective there is a need to standardize seismic specifications, identify a seismic specialist within each utility and enhance communications among these specialists. There is a general need to incorporate good seismic design practices on a national basis emphasizing new construction

  8. Seismic capacity of a reinforced concrete frame structure without seismic detailing and limited ductility seismic design in moderate seismicity

    International Nuclear Information System (INIS)

    Kim, J. K.; Kim, I. H.

    1999-01-01

    A four-story reinforced concrete frame building model is designed for the gravity loads only. Static nonlinear pushover analyses are performed in two orthogonal horizontal directions. The overall capacity curves are converted into ADRS spectra and compared with demand spectra. At several points the deformed shape, moment and shear distribution are calculated. Based on these results limited ductility seismic design concept is proposed as an alternative seismic design approach in moderate seismicity resign

  9. Seismic re-evaluation of Mochovce nuclear power plant. Seismic reevaluation of civil structures

    International Nuclear Information System (INIS)

    Podrouzek, P.

    1997-01-01

    In this contribution, an overview of seismic design procedures used for reassessment of seismic safety of civil structures at the Mochovce NPP in Slovak Republic presented. As an introduction, the objectives, history, and current status of seismic design of the NPP have been explained. General philosophy of design methods, seismic classification of buildings, seismic data, calculation methods, assumptions on structural behavior under seismic loading and reliability assessment were described in detail in the subsequent section. Examples of calculation models used for dynamic calculations of seismic response are given in the last section. (author)

  10. Seismic design standardization of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.; Vaze, K.K.

    2011-01-01

    Full text: Structures, Systems and Components (SSCs) of Nuclear Facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Man made accidents such as aircraft impact, explosions etc., some times may be considered as design basis event and some times taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event. It is generally felt design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to be adopted for seismic design standardization of nuclear facilities

  11. The seismic response and floor spectra of OL3 NPP buildings in Finland

    International Nuclear Information System (INIS)

    Pentti Varpasuo

    2005-01-01

    The purpose of the present work is the computation of seismic response and floor spectra of the nuclear power plant OL3 buildings in Olkiluoto. The following OL3 plant buildings were included in the analysis: 1. the Reactor Building UJA/UJB; 2. the Safeguard Buildings UJH/UJK 1-4; 3. and the Fuel Building UFA The in-structure spectra were generated using the ground motion response spectra documented in YVL GUIDE 2.6 'Seismic events at nuclear power plants' issued by Finnish Centre of Radiation Protection. The floor spectra were computed for the following equipment damping values: 2%, 4%, 7%, and 10%. The joint model for the plant buildings was generated. All analyses were linear and the direct time integration method was used with time step of 0.001 sec. All response runs were carried out with MSC/Nastran general purpose structural analysis program. The development of floor spectra has been carried out in accordance with the US NRC -Regulatory Guide 1.122: 'Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components'. The response results show that the dominant frequencies of the reactor building are located around 5 Hz in frequency space and that the typical amplification of spectral peaks for 4% damping is from 8 -10 times when compared to peak ground acceleration. (authors)

  12. Seismic-design questions typify nuclear obstacles

    International Nuclear Information System (INIS)

    Strauss, S.D.

    1979-01-01

    The trade-off between safe design of nuclear power plants and cost is considered. As an example, seismic protection problems at the Beaver Valley station of Duquesne Light Co. and their resolution by Stone and Webster Engineering are discussed

  13. Evaluation of high frequency ground motion effects on the seismic capacity of NPP equipments

    International Nuclear Information System (INIS)

    Choi, In Kil; Seo, Jeong Moon; Choun, Young Sun

    2003-04-01

    In this study, the uniform hazard spectrum for the example Korean nuclear power plants sites were developed and compared with various response spectra used in past seismic PRA and SMA. It shows that the high frequency ground motion effects should be considered in seismic safety evaluations. The floor response spectra were developed using the direct generation method that can develop the floor response spectra from the input response spectrum directly with only the dynamic properties of structures obtained from the design calculation. Most attachment of the equipments to the structure has a minimum distortion capacity. This makes it possible to drop the effective frequency of equipment to low frequency before it is severely damaged. The results of this study show that the high frequency ground motion effects on the floor response spectra were significant, and the effects should be considered in the SPRA and SMA for the equipments installed in a building. The high frequency ground motion effects are more important for the seismic capacity evaluation of functional failure modes. The high frequency ground motion effects on the structural failure of equipments that attached to the floor by welding can be reduced by the distortion capacity of welded anchorage

  14. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M; Boncheva, H; Stafanov, D [Central Laboratory for Seismic Mechanics and Earthquake Engineering, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    1993-07-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation.

  15. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    International Nuclear Information System (INIS)

    Kostov, M.; Boncheva, H.; Stafanov, D.

    1993-01-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation

  16. Seismic upgrading of WWER 440-230 structures, units 1/2, Kozloduy NPP

    International Nuclear Information System (INIS)

    Stafanov, D.; Kostov, M.; Boncheva, H.; Varbanov, G.

    1995-01-01

    The purpose of this paper is to present final results from a big amount of computational work in connection with the investigations of the possibilities for upgrading of WWER 440-230 structures, units 1/2, Kozloduy NPP. (author)

  17. Seismic response and resistance capacity of 'as built' WWER 440-230 NPP Kozloduy: Verification of the results by experiments and real earthquake

    International Nuclear Information System (INIS)

    Sachanski, S.

    1993-01-01

    Although Kozloduy NPP units 1 and 2 were not designed for earthquakes they have withstood successfully the Vrancea Earthquake in 1977 with sire peak ground acceleration of 83 sm/s 2 . Both units as well as units 3 and 4 were later recalculated for maximum peak acceleration of 0.1 g. According to values calculated by two-dimensional model, in 1980 reactor buildings had sufficient earthquake resistance capacity for the accepted design seismic excitation. The non symmetric design of WWER-440 structures in plan and elevation, the large eccentricity between the center of rigidities and masses as well as technological connections between the separate substructures and units led to complicated space response and rotational effects which cannot be calculated by two-dimensional models. Three dimensional detailed 'as built' mathematical models were established and verified by series of experiments and real earthquake for: detailed analysis of 'as built' structural response, comparing the results of two and three dimensional models, detailed analyses of seismic safety margins

  18. NRC Seismic Design Margins Program Plan

    International Nuclear Information System (INIS)

    Cummings, G.E.; Johnson, J.J.; Budnitz, R.J.

    1985-08-01

    Recent studies estimate that seismically induced core melt comes mainly from earthquakes in the peak ground acceleration range from 2 to 4 times the safe shutdown earthquake (SSE) acceleration used in plant design. However, from the licensing perspective of the US Nuclear Regulatory Commission, there is a continuing need for consideration of the inherent quantitative seismic margins because of, among other things, the changing perceptions of the seismic hazard. This paper discusses a Seismic Design Margins Program Plan, developed under the auspices of the US NRC, that provides the technical basis for assessing the significance of design margins in terms of overall plant safety. The Plan will also identify potential weaknesses that might have to be addressed, and will recommend technical methods for assessing margins at existing plants. For the purposes of this program, a general definition of seismic design margin is expressed in terms of how much larger that the design basis earthquake an earthquake must be to compromise plant safety. In this context, margin needs to be determined at the plant, system/function, structure, and component levels. 14 refs., 1 fig

  19. Main results of substantiation of the ecological safety of the Novovoronezh NPP-2 design

    International Nuclear Information System (INIS)

    Kopytov, I.I.; Kocher'yan, V.M.; Leonov, S.V.; Chionov, V.G.; Ehrnestova, L.S.

    2005-01-01

    Paper presents the results of the efforts to determine both the actual (hydrochemical, hydrobiological, geobotanical, soil, radiological) and the predicted parameters of the region ecology derived when substantiating the ecological safety of the Novovoronezh NPP-2 design [ru

  20. Russian seismic standards and demands for equipment and their conformity with international standards

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

    The principle regulations of standard documents concerning seismic safety of NPPs and demands for reactor equipment conformity with international standards are presented in this report. General state of NPP safety standards is reviewed, with a special emphasis on the state of seismic design standards for NPP equipment and piping. Russian standards documents on seismic resistance of NPPs and requirements are compared to international ones

  1. Seismic Design Guidelines For Port Structures

    DEFF Research Database (Denmark)

    Burcharth, H. F.; Bernal, Alberto; Blazquez, Rafael

    In order to mitigate hazards and losses due to earthquakes, seismic design methodologies have been developed and implemented in design practice in many regions since the early twentieth century, often in the form of codes and standards. Most of these methodologies are based on a force-balance app...

  2. Key issues in european reactor seismic design

    International Nuclear Information System (INIS)

    Cicognani, G.; Martelli, A.

    1984-01-01

    The paper focuses on the main problems which have arisen in FBR design in Europe due to seismic conditions. Its first part, derived from the final report of a CEC-Belgonucleaire study contract, clarifies how ''real'' is the seismic problem for each site. Then, the second and main part deals with the studies carried out in the european countries on the relevant subjects, typical of FBRs or related to specific needs of single FBRs: these studies, for which contributions were provided by ENEA, CEA, NNC and INTERATOM, concern mainly the numerical and experimental analysis of the core, the reactor vessel, the shut-down system and the reactor building of FBRs under construction or in advanced design phase. Attention is also paid to the studies started for future purposes, the feed-backs on the design due to seismic conditions, and the instructions for future reactors

  3. Seismic design and analysis methods

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1993-01-01

    Seismic load is in many areas of the world the most important loading situation from the point of view of structural strength. Taking this into account it is understandable, that there has been a strong allocation of resources in the seismic analysis during the past ten years. In this study there are three areas of the center of gravity: (1) Random vibrations; (2) Soil-structure interaction and (3) The methods for determining structural response. The solution of random vibration problems is clarified with the aid of applications in this study and from the point of view of mathematical treatment and mathematical formulations it is deemed sufficient to give the relevant sources. In the soil-structure interaction analysis the focus has been the significance of frequency dependent impedance functions. As a result it was obtained, that the description of the soil with the aid of frequency dependent impedance functions decreases the structural response and it is thus always the preferred method when compared to more conservative analysis types. From the methods to determine the C structural response the following four were tested: (1) The time history method; (2) The complex frequency-response method; (3) Response spectrum method and (4) The equivalent static force method. The time history appeared to be the most accurate method and the complex frequency-response method did have the widest area of application. (orig.). (14 refs., 35 figs.)

  4. Requirements and possible upgrading concept for the WWER-440/213: Mochovce NPP structures under seismic conditions

    International Nuclear Information System (INIS)

    Freiman, M.

    1993-01-01

    The Mochovce-Nuclear Power Plant is one of the WWER-440/213 plants which has been designed against earthquake. Nevertheless, the design earthquake has not been assessed adequately to the seismic hazard at the site. A new seismic design shall include an increased seismic input and assure an acceptable standard of safety. This contribution is related to some design aspects of civil structures for this nearly finished plant, such as: existing design and its margins with regard to the employed codes; requirements for a new design concept; effects to be expected by an increased design earthquake; applicable design methods; use of inelastic design spectra, behavior factors and capacity design; feasible upgrading measures. (author)

  5. Design experience on seismically isolated buildings

    International Nuclear Information System (INIS)

    Giuliani, G.C.

    1991-01-01

    This paper describes the practical problems associated with the structural design of seismically isolated buildings now under construction in Ancona, Italy. These structures are the first seismically isolated buildings in Italy. The Ancona region is in zone 2 of the Italian Seismic Code. It has a design acceleration of 0.07 g which corresponds to a ground surface acceleration of 0.25 g. The last significant earthquake was recorded on June 14, 1972, having a single shock-type wave with a peak acceleration of 0.53 g. Taking into account the aforesaid earthquake, the structural design of these new buildings was performed according to an acceleration spectrum which was different from the zone 2 seismic code and which provided protection for stronger ground motions. To minimize the cost of the structure, the buildings used ribbed plate decks, thus reducing the amount of material and the mass of the structures to be isolated. The design requirements, dynamic analysis performed, structural design, and practical engineering employed are reported in this paper. A comparison between the costs of a conventionally designed and a base-isolated structure is also reported. It shows a net savings of 7% for the base-isolated structure. The tests undertaken for certifying the mechanical properties of the isolators for both static and dynamic loads are also described, as is the full-scale dynamic test which is scheduled for next year (1990) for one of the completed buildings. (orig.)

  6. Design experience on seismically isolated buildings

    International Nuclear Information System (INIS)

    Giuliani, G.C.

    1989-01-01

    This paper describes the practical problems associated with the structural design of a group of seismically isolated buildings now under construction in Ancona, Italy. These structures are the first seismically isolated buildings in Italy. Taking into account previous earthquakes, the structural design of these new buildings was performed according to an acceleration spectrum which was different from its Zone 2 seismic code and which provided protection for stronger ground motions. To minimize the cost of the structure, the buildings used ribbed plate decks, thus reducing the amount of material and the mass of the structures to be isolated. The design requirements, dynamic analysis performed, structural design, and practical engineering employed are reported in this paper. A comparison between the costs of a conventionally designed and a base-isolated structure is also reported. The tests undertaken for certifying the mechanical properties of the isolators for both static and dynamic loads are also described, as is the full-scale dynamic test which is scheduled for next year (1990) for one of the completed buildings. Lessons learned in this design effort are potentially applicable to seismic base isolation for nuclear power plants

  7. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M.K. [Bulgarian Academy of Sciences, Sofia (BG). Central Lab. for Seismic Mechanics and Earthquake Engineering; Ma, D.C. [Argonne National Lab., IL (United States); Prato, C.A. [Univ. of Cordoba (AR); Stevenson, J.D. [Stevenson and Associates, Cleveland, OH (US)

    1993-08-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made.

  8. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    International Nuclear Information System (INIS)

    Kostov, M.K.; Prato, C.A.; Stevenson, J.D.

    1993-01-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made

  9. Domestic design and validation of natural circulation steam generator of China 1000 MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu, H.Y.; Wang, X.Y.; Wu, G.; Qin, J.M.; Xiong, Ch.H.; Wang, W.; Chen, J.L.; Cheng, H.P.; Zuo, Ch.P.

    2005-01-01

    In order to meet the requirements of domestic design of China intending built NPP projects, Research Institute of Nuclear Power Operation (RINPO) has achieved design of 1000 MWe NPP steam generator, called RINSG-1000(means 1000MWe SG designed by RINPO), which is based on SG research ,experiments and service experience accumulated by RINPO in more 40 years. Testing validation of two steam generator key technologies, advanced moisture separate device and sludge collector, has been accomplished during the period of 2000 to 2002. This paper describes the design features of RINSG-1000, and provides some validation test results. (authors)

  10. Reassessment of Seismic Design and Noise Simulation using Finite Element Calculation of the Condensate Storage Tank of Cofrentes NPP according to standard API-650 11th Ed; Reevaluacion del diseno Sismico mediante Simulacion de Fluidos y Calculo por Elementos Finitos del Deposito del Almacenamiento de Condensado de Central Nuclear de Cofrentes conforme a la norma API-650 11th Ed

    Energy Technology Data Exchange (ETDEWEB)

    Sarti Fernandez, F.; Gavilan Moreno, C.; Paez Ortega, E.

    2012-07-01

    There have been several dynamic simulations in which I analyzed: fluid-structure interaction effect of the wave, studying stress, vibration modes and possible effects of structural instability. After this process to make the changes in the tank to comply with the new rules and updated seismic conditions were designed. were performed.

  11. Conceptual design by analysis of KALIMER seismic isolation

    International Nuclear Information System (INIS)

    You, Bong; Koo, Kyung Hoi; Lee, Jae Han

    1996-06-01

    The objectives of this report are to preliminarily evaluate the seismic isolation performance of KALIMER (Korea Advance LIquid MEtal Reactor) by seismic analyses, investigate the design feasibility, and find the critical points of KALIMER reactor structures. The work scopes performed in this study are 1) the establishment of seismic design basis, 2) the development of seismic analysis model of KALIMER, 3) the modal analysis, 4) seismic time history analysis, 5) the evaluations of seismic isolation performance and seismic design margins, and 6) the evaluation of seismic capability of KALIMER. The horizontal fundamental frequency of KALIMER reactor structure is 8 Hz, which is far remote from the seismic isolation frequency, 0.7 Hz. The vertical first and second natural frequencies are about 2 Hz and 8 Hz respectively. These vertical natural frequencies are in a dominant ground motion frequency bands, therefore these modes will result in large vertical response amplifications. From the results of seismic time history analyses, the horizontal isolation performance is great but the large vertical amplifications are occurred in reactor structures. The RV Liner has the smallest seismic design margin as 0.18. From the results of seismic design margins evaluation, the critical design change are needed in the support barrel, separation plate, and baffle plate points. The seismic capability of KALIMER is about 0.35g. This value can be increased by the design changes of the separation plate and etc.. 11 tabs., 29 figs., 7 refs. (Author) .new

  12. Adapting standards to the site. Example of Seismic Base Isolation

    International Nuclear Information System (INIS)

    Viallet, Emmanuel

    2014-01-01

    Emmanuel Viallet, Civil Design Manager at EDF engineering center SEPTEN, concluded the morning's lectures with a presentation on how to adapt a standard design to site characteristics. He presented the example of the seismic isolation of the Cruas NPP for which the standard 900 MW design was indeed built on 'anti-seismic pads' to withstand local seismic load

  13. Initial data of seismic input and soil conditions of Kozloduy NPP site. Extension to Part 2 soil conditions, issued October '93

    International Nuclear Information System (INIS)

    Boyadjiev, Z.

    1995-01-01

    On the basis of the results of the carried out experimental (laboratory and in situ) investigations of the dynamic characteristics, the following conclusions for the Kozloduy NPP site are presented. (1) The established through experimental studies relationships for the shear module and the damping factor as strain dependent of representative samples of soils of the site profile, can be used for all similar soils in the profile in the different parts of the site, taking into account the possible differences by means of the initial shear module in the normalized relationship for the respective generalized soil type. (2) When solving the problems of the site response and the 'soil - structure analysis', the geotechnical seismic model of the 'free field' profile can be assumed for all parts of the NPP site. (3) The changes of the lithological profile in different parts of the site, in respect to type and thickness, as well as in view of the different way and depth of the NPP structures foundation, make it necessary the elaboration of a geotechnical seismic model of the profile below the foundation plates of the reactor buildings of the NPP units in each particular case. These models can be made out on the basis of the summarized data about the shear velocities of the soil types, the lithological data of the studied boreholes in these places and the data having natural bulk density from 30 - 40 m depth determined by the laboratory studies of samples of these soils, assuming with approximation that the geotechnical seismic model below this depth is the same as the one of the 'free field'. (4) Studies have been carried out through in situ and laboratory studies of all the fundamental structures on the NPP site and the results of them are sufficient as an addition to the present initial data for solving the problems of the site response and the 'soil-structure inter-action' analyses of each structure

  14. Salt Repository Project input to seismic design: Revision 0

    International Nuclear Information System (INIS)

    1987-12-01

    The Salt Repository Program (SRP) Input to Seismic Design (ISD) documents the assumptions, rationale, approaches, judgments, and analyses that support the development of seismic-specific data and information to be used for shaft design in accordance with the SRP Shaft Design Guide (SDG). The contents of this document are divided into four subject areas: (1) seismic assessment, (2) stratigraphy and material properties for seismic design, (3) development of seismic design parameters, and (4) host media stability. These four subject areas have been developed considering expected conditions at a proposed site in Deaf Smith County, Texas. The ISD should be used only in conjunction with seismic design of the exploratory and repository shafts. Seismic design considerations relating to surface facilities are not addressed in this document. 54 refs., 55 figs., 18 tabs

  15. Seismic qualification tests of fans of the NPP of Laguna Verde U-1 and U-2; Pruebas de calificacion sismica de ventiladores de la Central Laguna Verde U1 and U2

    Energy Technology Data Exchange (ETDEWEB)

    Jarvio C, G.; Garcia H, E. E.; Arguelles F, R.; Vela H, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Naranjo U, J. L., E-mail: gilberto.jarvio@inin.gob.mx [Comision Federal de Electricidad, Gerencia de Centrales Nucleoelectricas, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km 7.5, Dos Bocas, Veracruz (Mexico)

    2013-10-15

    This work presents the results of the seismic qualification tests applied to the fans that will be installed in the control panels of the three divisions of the diesel generators of the nuclear power plant (NPP) of Laguna Verde, Unit-1 and Unit-2. This seismic qualification process of the fans was carried out using two specimens that were tested in the seismic table (vibrating) of the Engineering Institute of Universidad Nacional Autonoma de Mexico (UNAM), in accordance with the requirements of the standard IEEE 344-1975, to satisfy the established requirements of seismic qualification in the technical specifications and normative documents required by the nuclear standards, in order to demonstrate its application in the diesel generators Divisions I, II and III of the NPP. The seismic qualification tests were developed on specimens that were retired of the NPP of Laguna Verde recently with a service life of 7.75 years. (Author)

  16. International cooperation in the field of studying seismic resistance of NPP components

    International Nuclear Information System (INIS)

    Kaznovskij, S.P.; Chechenov, Kh.D.

    1989-01-01

    Main results of the conference of representations from the USSR, Bulgarie, Hungary and Chechoslovakia related to the problems of seismology and seismic resistance of NPPs are briefly formulated. One of the important results of the conference consists in the agrement concerning cooperation and mutual application of seismoexplosive testing ground near Nalchik

  17. Key features of MIR.1200 (AES-2006) design and current stage of Leningrad NPP-2 construction

    International Nuclear Information System (INIS)

    Ivkov, Igor

    2010-01-01

    MIR.1200/AES-2006 is an abbreviated name of the evolving NPP design developed on the basis of the VVER-1000 Russian design with gross operation life of 480 reactor-years. This design is being implemented in four Units of Leningrad NPP-2 (LNPP-2. The AES-91/99 was used as reference during development of the AES-2006 design for LNPP-2; this design was implemented in two Units of Tianwan NPP (China). The main technical features of the MIR.1200/AES-2006 design include a double containment, four trains of active safety systems (4x100%, 4x50%), and special engineering measures for BDBA management (core catcher, H2 PARs, PHRS) based mainly on passive principles. The containment is described in detail, the main features in comparison with the reference NPP are outlined, the design layout principles are highlighted, the safety system structure and parameters are described. Attention is paid to the BDBA management system, hydrogen removal system, core catcher, and PHRS-SG and C-PHRS. (P.A.)

  18. NPP Design Basis Handover and Knowledge Preservation from Subcontractors, Vendors and EPC

    International Nuclear Information System (INIS)

    Freeland, Kent

    2013-01-01

    Using PLM-based Workflow for Configuration Management (CM) in the Nuclear Power Industry Advantages – some work to do! • NPP’s must adapt to using PLM-based solutions to support CM and to synchronize design changes to asset or product changes, and reduce “slipstreaming”. In the NPP world, this often appears as events that circumvent CM – for example, non-approved parts substitutions and “temporary” plant modifications that are never removed. • PLM serves as the method for unifying the application of requirements to design changes, processes and workflow. In NPP’s, requirements are generally considered only relevant to designs – not process and workflow. • PLM supports Configuration Management and Design Basis in Regulator Action Tracking for NPP’s, and application of PLM-based CM to regulator action and compliance systems. This is a poorly-understood application of CM in NPP’s, yet these elements control large parts of the NPP design basis. • Suppliers, EPC’s and Technology Vendors must also understand the role of CM, SE and PLM in construction of new standards-driven NPP designs (like EPR and Westinghouse AP-1000 NPP designs), as well as understanding the role and handling of Knowledge Systems

  19. Design and implementation experience of seismic upgrades at Kozloduy and Paks NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Borov, V; Trichkov, V; Alexandrov, A; Jordanov, M [EQE-Bulgaria, Sofia (Bulgaria)

    1995-07-01

    Series of upgrades have been designed and implemented by EQE-Bulgaria at Kozloduy NPP and as a subcontractor of EQE-International - at Paks NPP. Wide variety of facilities have been upgraded, including Electrical Equipment, Control and Instrumentation Equipment, Technological Equipment, Brick Walls and Building Structures. Different design approaches and concepts have been applied in compliance with the specific technological and structural conditions. The effect of the excitation intensity as well as the presence of specific floor response spectra over the upgrading concept and cost is discussed. Specific problems of supporting heavy technological equipment are noted. A practical approach for seismic upgrading of Brick Walls, as well as a tendency for unification of the engineering design is shown. The first completely upgraded Building Structure at Kozloduy NPP is the structure of the Electrical Control Building to the Diesel Generator of the River-bank Pump Station. Specific problems of the implementation of the final upgrading design of the Diesel Generator Building are outlined. (author)

  20. Design and implementation experience of seismic upgrades at Kozloduy and Paks NPPs

    International Nuclear Information System (INIS)

    Borov, V.; Trichkov, V.; Alexandrov, A.; Jordanov, M.

    1995-01-01

    Series of upgrades have been designed and implemented by EQE-Bulgaria at Kozloduy NPP and as a subcontractor of EQE-International - at Paks NPP. Wide variety of facilities have been upgraded, including Electrical Equipment, Control and Instrumentation Equipment, Technological Equipment, Brick Walls and Building Structures. Different design approaches and concepts have been applied in compliance with the specific technological and structural conditions. The effect of the excitation intensity as well as the presence of specific floor response spectra over the upgrading concept and cost is discussed. Specific problems of supporting heavy technological equipment are noted. A practical approach for seismic upgrading of Brick Walls, as well as a tendency for unification of the engineering design is shown. The first completely upgraded Building Structure at Kozloduy NPP is the structure of the Electrical Control Building to the Diesel Generator of the River-bank Pump Station. Specific problems of the implementation of the final upgrading design of the Diesel Generator Building are outlined. (author)

  1. A procedure for the determination of scenario earthquakes for seismic design based on probabilistic seismic hazard analysis

    International Nuclear Information System (INIS)

    Hirose, Jiro; Muramatsu, Ken

    2002-03-01

    This report presents a study on the procedures for the determination of scenario earthquakes for seismic design of nuclear power plants (NPPs) based on probabilistic seismic hazard analysis (PSHA). In the recent years, the use of PSHA, which is a part of seismic probabilistic safety assessment (PSA), to determine the design basis earthquake motions for NPPs has been proposed. The identified earthquakes are called probability-based scenario earthquakes (PBSEs). The concept of PBSEs originates both from the study of US NRC and from Ishikawa and Kameda. The assessment of PBSEs is composed of seismic hazard analysis and identification of dominant earthquakes. The objectives of this study are to formulate the concept of PBSEs and to examine the procedures for determining the PBSEs for a domestic NPP site. This report consists of three parts, namely, procedures to compile analytical conditions for PBSEs, an assessment to identify PBSEs for a model site using the Ishikawa's concept and the examination of uncertainties involved in analytical conditions. The results obtained from the examination of PBSEs using Ishikawa's concept are as follows. (a) Since PBSEs are expressed by hazard-consistent magnitude and distance in terms of a prescribed reference probability, it is easy to obtain a concrete image of earthquakes that determine the ground response spectrum to be considered in the design of NPPs. (b) Source contribution factors provide the information on the importance of the earthquake source regions and/or active faults, and allows the selection of a couple of PBSEs based on their importance to the site. (c) Since analytical conditions involve uncertainty, sensitivity analyses on uncertainties that would affect seismic hazard curves and identification of PBSEs were performed on various aspects and provided useful insights for assessment of PBSEs. A result from this sensitivity analysis was that, although the difference in selection of attenuation equations led to a

  2. Research on performance-based seismic design criteria

    Institute of Scientific and Technical Information of China (English)

    谢礼立; 马玉宏

    2002-01-01

    The seismic design criterion adopted in the existing seismic design codes is reviewed. It is pointed out that the presently used seismic design criterion is not satisfied with the requirements of nowadays social and economic development. A new performance-based seismic design criterion that is composed of three components is presented in this paper. It can not only effectively control the economic losses and casualty, but also ensure the building(s function in proper operation during earthquakes. The three components are: classification of seismic design for buildings, determination of seismic design intensity and/or seismic design ground motion for controlling seismic economic losses and casualties, and determination of the importance factors in terms of service periods of buildings. For controlling the seismic human losses, the idea of socially acceptable casualty level is presented and the (Optimal Economic Decision Model( and (Optimal Safe Decision Model( are established. Finally, a new method is recommended for calculating the importance factors of structures by adjusting structures service period on the base of more important structure with longer service period than the conventional ones. Therefore, the more important structure with longer service periods will be designed for higher seismic loads, in case the exceedance probability of seismic hazard in different service period is same.

  3. End of mission report on seismic safety review mission for Belene NPP site

    International Nuclear Information System (INIS)

    Gurpinar, A.; Mohammadioun, B.; Schneider, H.; Serva, L.

    1995-01-01

    Upon the invitation of the Bulgarian government through the Committee for the Peaceful Uses of Atomic Energy and within the framework of the implementation of the Technical Cooperation project BUL/9/012 related to site and seismic of NPPs, a mission visited Sofia 3 - 7 July 1995. The mission constituted a follow-up of the interim review of subjects related to tectonic stability and seismic hazard characterization of the site which was performed in September 1993. The main objective of the mission was the final review of the subjects already reviewed in September 1993 as well as issues related to geotechnical engineering and foundation safety. The main terms of reference of the present mission was to verify the implementation of the recommendations of the Site Safety Review Mission of June 1990. This document gives findings on geology-tectonics, seismology and foundation safety. In the end conclusions and recommendations of the mission are presented

  4. Design of NPP of new generation being constructed at Novovoronezh NPP site

    International Nuclear Information System (INIS)

    Afrov, A.; Berkovich, V.; Generalov, V.; Dragunov, Yu.; Krushelnitskij, V.

    2000-01-01

    The design of a new-generation nuclear power plant with advanced WWER-1000 units, currently under construction at Novovoronezh, is dealt with in considerable detail. Information is given on the general layout of the power plant, and the monoblock comprising the four-loop reactor plant with double containment, the turbine hall with two turbine-driven feedwater pumps and safety and auxiliary systems is described. In a separate building at the power plant site, a new fuel storage facility is located designed for extreme external effects. The operations involved in fuel handling, in both the storage facility and the reactor department, are mentioned. The management of various types of radioactive waste is highlighted. The basic principles of the engineering solutions pertaining to the plant's electrical part are outlined. The concept of the instrumentation and control system is explained. Information is also given on the double containment system. A considerable part of the presentation is devoted to the philosophy and concept of the power plant safety, to the results of safety evaluation and to the research carried out in this respect. The paper is concluded with a discussion of the experiments and analyses performed with the aim to justify the design solutions as regards the passive heat removal system. (A.K.)

  5. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP.

  6. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP

  7. Development of Canadian seismic design approach and overview of seismic standards

    Energy Technology Data Exchange (ETDEWEB)

    Usmani, A. [Amec Foster Wheeler, Toronto, ON (Canada); Aziz, T. [TSAziz Consulting Inc., Mississauga, ON (Canada)

    2015-07-01

    Historically the Canadian seismic design approaches have evolved for CANDU® nuclear power plants to ensure that they are designed to withstand a design basis earthquake (DBE) and have margins to meet the safety requirements of beyond DBE (BDBE). While the Canadian approach differs from others, it is comparable and in some cases more conservative. The seismic requirements are captured in five CSA nuclear standards which are kept up to date and incorporate lessons learnt from recent seismic events. This paper describes the evolution of Canadian approach, comparison with others and provides an overview and salient features of CSA seismic standards. (author)

  8. Key developments in the advanced NPP with WWER-640/V-407 reactor plant design

    International Nuclear Information System (INIS)

    Dragunov, Yu.G.; Mokhov, V.A.; Nikitenko, M.P.; Afrov, A.M.

    1999-01-01

    The report covers the main design features of advanced NPP equipped with WWER-640 reactor, that take into account the up-to-date approaches in the process of forming safety concepts. An approach to accident management has been analysed, beyond design-basis accidents included. A description of principal safety systems has been presented as well as the interrelation of their operation. The principal features of the systems design have been shown. (author)

  9. Response of a NPP reactor building under seismic action with regard to different soil properties

    International Nuclear Information System (INIS)

    Wagenknecht, E.

    1987-01-01

    The object of this investigation is the response of a reactor building on seismic action with systematic variation of the soil stiffness. A thin-walled orthotropic containment shell on varying heavy and rigid foundations is regarded as calculation model. The soil stiffness is simulated by meand of spring elements for horizontal translation and for rocking motions of the building. By the response spectra method the loads of the containment shell are calculated for a horizontal seismic excitation. The investigation is aimed at determining the influence of differentiated soil stiffnesses on the containment action effects and at recognizing the causes for the occuring effects. The results are thoroughly represented by selected quantities of the building's response, the effects from the soil-structure interaction are discussed and the causes of the effects cleary explained. Apossibility is provided for determining critical soil stiffnesses which cause a siginificat intensification effect. The results of the investigations show that both the soil stiffness and structural configuration of the reactor building particulary in case of the substructure being heavy and rigid, exert a decisive on the loading of the superstructure. (orig.)

  10. Design and analysis of CANDU NPP internal structures for Japanese conditions

    International Nuclear Information System (INIS)

    Aziz, T.S.; Murakami, H.

    1991-01-01

    The design and analysis approach for the CANDU 6 Internal Concrete Structure (ICS) for Japanese seismic conditions is described. The approach consists of a seismic analysis to determine the design level accelerations; followed by a detailed finite element analysis to determine the section forces for each shell element. The extent of the design modifications for the original structure to meet the Japanese design conditions is given. (author)

  11. Design Basis Threat (DBT) Approach for the First NPP Security System in Indonesia

    International Nuclear Information System (INIS)

    Ign Djoko Irianto

    2004-01-01

    Design Basis Threat (DBT) is one of the main factors to be taken into account in the design of physical protection system of nuclear facility. In accordance with IAEA's recommendations outlined in INFCIRC/225/Rev.4 (Corrected), DBT is defined as: attributes and characteristics of potential insider and/or external adversaries, who might attempt unauthorized removal of nuclear material or sabotage against the nuclear facilities. There are three types of adversary that must be considered in DBT, such as adversary who comes from the outside (external adversary), adversary who comes from the inside (internal adversary), and adversary who comes from outside and colludes with insiders. Current situation in Indonesia, where many bomb attacks occurred, requires serious attention on DBT in the physical protection design of NPP which is to be built in Indonesia. This paper is intended to describe the methodology on how to create and implement a Design Basis Threat in the design process of NPP physical protection in Indonesia. (author)

  12. Seismic verification of the Italian PEC fast reactor and effects of seismic conditions on the design

    International Nuclear Information System (INIS)

    Martelli, A.; Cecchini, F.; Masoni, P.; Maresca, G.; Castoldi, A.

    1988-01-01

    This paper deals with the aseismic design features of the Italian PEC fast reactor and the effects of seismic conditions on the reactor design. More precisely, after some notes on the main plant features, the paper reports on the design earthquakes adopted, the seismic monitoring procedures and the related actions, the design requirements, criteria and methods, and also provides a brief summary of the main research and development studies performed in support of design analysis. For the above-mentioned items, comparisons with the other fast reactors of the European Community countries are presented. Furthermore, the paper stresses the design modifications adopted to guarantee PEC seismic safety

  13. Comparative study for methods to determine the seismic response of NPP structures

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1995-01-01

    There are many different important problem areas in evaluating the seismic response of structures. In this study the effort is concentrated on three of these areas. The first task is the mathematical formulation of earthquake excitation. The random vibration theory is taken as the tool in this task. The second area of interest in this study is the soil-structure interaction analysis. The approach of impedance functions is chosen and the focal point of interest is the significance of frequency dependent impedance functions. The third area of interest is the methods to determine the structural response. The following three methods were tested: the mode superposition time history method; the complex frequency response method; the response spectrum method. The comparison was made with the aid of MSC/NASTRAN code. The three methods gave for outer containment building response results which were in good agreement with each other. (author). 4 refs., 5 figs

  14. Comparative study of codes for the seismic design of structures

    Directory of Open Access Journals (Sweden)

    S. H. C. Santos

    Full Text Available A general evaluation of some points of the South American seismic codes is presented herein, comparing them among themselves and with the American Standard ASCE/SEI 7/10 and with the European Standard Eurocode 8. The study is focused in design criteria for buildings. The Western border of South America is one of the most seismically active regions of the World. It corresponds to the confluence of the South American and Nazca plates. This region corresponds roughly to the vicinity of the Andes Mountains. This seismicity diminishes in the direction of the comparatively seismically quieter Eastern South American areas. The South American countries located in its Western Border possess standards for seismic design since some decades ago, being the Brazilian Standard for seismic design only recently published. This study is focused in some critical topics: definition of the recurrence periods for establishing the seismic input; definition of the seismic zonation and design ground motion values; definition of the shape of the design response spectra; consideration of soil amplification, soil liquefaction and soil-structure interaction; classification of the structures in different importance levels; definition of the seismic force-resisting systems and respective response modification coefficients; consideration of structural irregularities and definition of the allowable procedures for the seismic analyses. A simple building structure is analyzed considering the criteria of the several standards and obtained results are compared.

  15. Safety design guides for seismic requirements for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for seismic requirements for CANDU 9 describes the seismic design philosophy, defines the applicable earthquakes and identifies the structures and systems requiring seismic qualification to ensure that the essential safety function can be adequately satisfied following earthquake. The detailed requirements for structures, systems and components which must be seismically qualified are specified in the Appendix. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 fig., (Author) .new

  16. Structural Analysis and Seismic Design for Cold Neutron Laboratory Building

    International Nuclear Information System (INIS)

    Wu, Sangik; Kim, Y. K.; Kim, H. R.

    2007-05-01

    This report describes all the major results of the dynamic structural analysis and seismic design for the Cold Neutron Laboratory Building which is classified in seismic class II. The results are summarized of the ground response spectrum as seismic input loads, mechanical properties of subsoil, the buoyancy stability due to ground water, the maximum displacement of the main frame under the seismic load and the member design. This report will be used as a basic design report to maintenance its structural integrity in future

  17. Building configuration and seismic design: The architecture of earthquake resistance

    Science.gov (United States)

    Arnold, C.; Reitherman, R.; Whitaker, D.

    1981-05-01

    The architecture of a building in relation to its ability to withstand earthquakes was determined. Aspects of round motion which are significant to building behavior are discussed. Results of a survey of configuration decisions that affect the performance of buildings with a focus on the architectural aspects of configuration design are provided. Configuration derivation, building type as it relates to seismic design, and seismic design, and seismic issues in the design process are examined. Case studies of the Veterans' Administration Hospital in Loma Linda, California, and the Imperial Hotel in Tokyo, Japan, are presented. The seismic design process is described paying special attention to the configuration issues. The need is stressed for guidelines, codes, and regulations to ensure design solutions that respect and balance the full range of architectural, engineering, and material influences on seismic hazards.

  18. Assessment of seismic design response factors of concrete wall buildings

    Science.gov (United States)

    Mwafy, Aman

    2011-03-01

    To verify the seismic design response factors of high-rise buildings, five reference structures, varying in height from 20- to 60-stories, were selected and designed according to modern design codes to represent a wide range of concrete wall structures. Verified fiber-based analytical models for inelastic simulation were developed, considering the geometric nonlinearity and material inelasticity of the structural members. The ground motion uncertainty was accounted for by employing 20 earthquake records representing two seismic scenarios, consistent with the latest understanding of the tectonic setting and seismicity of the selected reference region (UAE). A large number of Inelastic Pushover Analyses (IPAs) and Incremental Dynamic Collapse Analyses (IDCAs) were deployed for the reference structures to estimate the seismic design response factors. It is concluded that the factors adopted by the design code are adequately conservative. The results of this systematic assessment of seismic design response factors apply to a wide variety of contemporary concrete wall buildings with various characteristics.

  19. Structure analysis and design of PCCV for new generation NPP

    International Nuclear Information System (INIS)

    Wang Mingdan; Wang Xiaowen; Huang Xiaolin; Xia Zufeng

    2005-01-01

    The paper documents the overall schedule work which has been done by Shanghai Nuclear Engineering Research and Design Institute (SNERDI) in the research and design scope of the new generational advanced prestressed concrete containment vessel (PCCV). It can be applied to the design of nuclear engineering and general prestressed concrete structures in civil engineering. (authors)

  20. Condenser Design for the Proposed AM600 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Md. Mizanur; Abdallah, Khaled Atya Ahmed; Field, Robert M. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The design goals are to make the condenser more robust and compact with a reduced component count. The AM600 condenser design also has new features as described below. Considering that the minimum heat sink temperature for potentially emergent nuclear countries is on the order of 21.deg. C or higher, a turbine design with a single low pressure rotor can be considered without sacrificing thermal efficiency. The condenser back pressure range for the considered markets is on the order of 2 to 3 in-HgA. With these boundary conditions, the AM600 condenser duty can be met with a single pressure zone design with a total of eight (8) titanium tube bundles (four (4) per pass) divided into four isolable sections. Due to the compact design (i.e., accepting exhaust from only one low pressure cylinder), both axial ends of the condenser are unobstructed and available for attachment of extended flash chambers, diverting inflows away from the tube bundles. The single shell design of this condenser then allows for an innovative design feature, namely the extended flash chambers. This permits the routing of dump, drain, vent, and bypass flows directly to these chambers, bypassing the condenser shell. Within the condenser shell, this design eliminates impingement plates, impingement boxes, and spargers. Failure of these components represents an ongoing source of condenser tube damage in operating nuclear units, requiring significant resources for outage inspections. The extended flash chamber approach also has a number of other advantages as delineated above.

  1. Seismic analysis for conceptual design of HCCR TBM-set

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won, E-mail: dwlee@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Kim, Suk-Kwon; Yoon, Jae Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon, Republic of Korea (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The seismic analysis of KO HCCR TBM-set are performed. • The seismic envents like SL-1, SL-2, and SMHV are selected and evaluated with FEM code (ANSYS). • The results of the stresses and deformations are confirmed to meet the design criteria. - Abstract: Using the conceptual design of the Korean helium cooled ceramic reflector (HCCR) test blanket module (TBM) including the TBM-shield for testing in ITER, a seismic analysis is performed. According to the ITER TBM port plug (TBM PP) system load specifications, seismic events are selected as SL-1 (seismic level-1), SL-2 (seismic level-2), and SMHV (seismes maximaux historiquement vraisemblables, Maximum Histroically Probable Earthquakes). In a modal analysis a total of 50 modes are obtained. Then, a spectra response analysis for each seismic event is carried out using ANSYS based on the modal analysis results. For each event, the obtained Tresca stress is evaluated to confirm the design integrity, by comparing the resulting stress to the design criteria. The Tresca strain and displacement are also estimated for the HCCR TBM-set. From the analysis, it was concluded that the maximum stresses by the seismic events meet the design criteria, and the displacements are lower than the designed gap from the TBM PP frame. The results are provided to a load combination analysis.

  2. Seismic design criteria for nuclear powerplants

    International Nuclear Information System (INIS)

    Jennings, P.C.; Guzman, R.A.

    1975-01-01

    There are three main aspects of the problem of selection of seismic design criteria for major projects such as nuclear power plants. These are the description of the appropriate level of shaking to be considered, usually given in the form of design spectra; the allowable response of the structure, usually specified in terms of allowable stresses and deflections; and the capability of the structure to dissipate energy, commonly given in the form of fractions of critical damping. In this presentation only the first of these features is examined, with particular application to nuclear power plants. Under these restrictions, the most important parts of the problem become the determination of the amplitude of the design spectra corresponding to the safe shutdown earthquake (SSE) and the question of whether the shape of the spectra recommended by Regulatory Guide 1.60 (U. S. Atomic Energy Commission, 1973) is appropriate for the particular application. In the course of working out the details of the approach, it was found useful to reexamine a number of concepts including the use of response spectra or peak values of ground motion parameters, the shape of the design spectra, problems in attenuation and scaling, and the use of motions on the ground surface or bedrock motions. There is nothing fundamentally new in the suggested approach, although some of the features may not have been applied to the problem of selecting design spectra for nuclear power plants in the way suggested. The approach is applied only to nuclear power plants but it is not limited to this application

  3. An intelligent and integrated V and V environment design for NPP I and C software systems

    International Nuclear Information System (INIS)

    Koo, Seo Ryong; Son Han Seong; Seong, Poong Hyun

    2001-01-01

    Nuclear Power Plant (NPP) is the safety critical system. Since, nuclear instrumentation and control (I and C) systems including the plant protection system play the brain part of human, nuclear I and C systems have an influence on safety and operation of NPP. Essentially, software V and V should be performed for the safety critical systems based on software. It is very important in the technical aspect because of the problems concerning license acquisitions. In this work, an intelligent and integrated V and V environment supporting the automation of V and V was designed. The intelligent and integrated V and V environment consists of the intelligent controller part, components part, interface part, and GUI part. These parts were integrated systematically, while taking their own independent functions

  4. Validation of the CORD-2 System for the Nuclear Design Calculations of the NPP Krsko Core

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

    The CORD-2 package intended for core design calculations of PWRs has be recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krsko core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krsko core.(author).

  5. Seismic Design of a Single Bored Tunnel: Longitudinal Deformations and Seismic Joints

    Science.gov (United States)

    Oh, J.; Moon, T.

    2018-03-01

    The large diameter bored tunnel passing through rock and alluvial deposits subjected to seismic loading is analyzed for estimating longitudinal deformations and member forces on the segmental tunnel liners. The project site has challenges including high hydrostatic pressure, variable ground profile and high seismic loading. To ensure the safety of segmental tunnel liner from the seismic demands, the performance-based two-level design earthquake approach, Functional Evaluation Earthquake and Safety Evaluation Earthquake, has been adopted. The longitudinal tunnel and ground response seismic analyses are performed using a three-dimensional quasi-static linear elastic and nonlinear elastic discrete beam-spring elements to represent segmental liner and ground spring, respectively. Three components (longitudinal, transverse and vertical) of free-field ground displacement-time histories evaluated from site response analyses considering wave passage effects have been applied at the end support of the strain-compatible ground springs. The result of the longitudinal seismic analyses suggests that seismic joint for the mitigation measure requiring the design deflection capacity of 5-7.5 cm is to be furnished at the transition zone between hard and soft ground condition where the maximum member forces on the segmental liner (i.e., axial, shear forces and bending moments) are induced. The paper illustrates how detailed numerical analyses can be practically applied to evaluate the axial and curvature deformations along the tunnel alignment under difficult ground conditions and to provide the seismic joints at proper locations to effectively reduce the seismic demands below the allowable levels.

  6. Design and approval of EPS diesel systems in German NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kollmer, A.A., E-mail: anton.kollmer@tuev-sued.de [TUV SUD Industrie Service GmbH, Munich (Germany)

    2014-07-01

    Not at least because of 'Fukushima accident', Emergency power supply (EPS) systems with diesel engines are regarded as higher safety-important as 20 years ago. This presentation shows the design and approval of the emergency power facilities of the German nuclear power plants. It deals with the essential details e. g.: EPS Diesel integration to main power grid of the German BWR and PWR; Procedure for 'Loss of off-site power (LOOP)-Design'; Design robustness of AC Power Supply (Design requirements: independent redundancies, airplane crash, explosion pressure wave, earth-quake); Cooling systems (chilled by river water, air, well water, stored water volume); Layout of the stationary emergency Diesel generator (fuel supply, starting system); Layout of the bunkered EDG with directly driven emergency feed water pump; Robustness of AC power supply beyond design; Layout of mobile chilling equipment for diesel engines; Layout of mobile Diesel gen-sets (200 kVA and 1250 kVA); German Requirements of KTA 3702 versus US-IEE 323; Construction, Materials and Testing of EDG; Maintenance (in-service inspections, operation, Repair); Lessons learned (e. g. crank house cracks, start failure due to too much oil in the combustion chamber) (author)

  7. Seismic analysis of a NPP reactor building using spectrum-compatible power spectral density functions

    International Nuclear Information System (INIS)

    Venancio Filho, F.; Joia, L.A.

    1987-01-01

    A numerical methodology to obtain Power Spectral Density Functions (PSDF) of ground accelerations, compatible with a given design response spectrum is presented. The PSDF's are derived from the statistical analysis of the amplitudes of the frequency components in a set of artificially generated time-histories matching the given spectrum. A so obtained PSDF is then used in the stochastic analysis of a reactor building. The main results of this analysis are compared with the ones obtained by deterministic methods. (orig./HP)

  8. Design features and operation experience of the main circulating pumps for the ''Loviisa'' NPP with the WWER-440 reactor

    International Nuclear Information System (INIS)

    Iofs, D.; Kujyala, I.; Timperi, I.; Shlejfer, G.; Vistbakka, V.; Prudovskij, A.M.; Turetskij, L.I.; Vorona, P.N.

    1980-01-01

    Technical characteristics and the operation of main circulating pumps (MCP) designed and mounted at the ''Loviisa'' NPP by Finnish firms ''Alstrem'' and ''Stremberg'' are described. The above MCP have specific advantages over similar pumps mounted at other NPP with pressurized water cooled reactors. This is a possibility of substitution of potentially most damaged units (bearing and pump shaft sealing) for several hours, without MCP disassembly as a whole as well as using rolling bearings together with the original electromagnetic unloading system from the axial force instead of usually employed in similar MCP radial thrust slip bearings. The two year operation experience has confirmed the efficiency and reliability of ''Loviisa'' NPP main circulating pumps

  9. Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

    International Nuclear Information System (INIS)

    Ambrosini, Daniel; Codina, Ramón H.; Curadelli, Oscar; Martínez, Carlos A.

    2017-01-01

    Highlights: • Structural analysis of CAREM-25 NPP is presented. • Full 3D numerical model was developed. • Transient thermal and static structural analyses were performed. • Modeling guidelines for numerical structural analysis of NPP are recommended. • Envelope condition of DBA dominates the structural behavior. - Abstract: In this paper, a numerical study about the structural response of the Argentine nuclear power plant CAREM-25 subjected to the design basis accident (DBA) and seismic loads is presented. Taking into account the hardware capabilities available, a full 3D finite element model was adopted. A significant part of the building was modeled using more than 2 M solid elements. In order to take into account the foundation flexibility, linear springs were used. The springs and the model were calibrated against a greater model used to study the soil-structure interaction. The structure was subjected to the DBA and seismic loads as combinations defined by ASME international code. First, a transient thermal analysis was performed with the conditions defined by DBA and evaluating the time history of the temperature of the model, each 1 h until 36 h. The final results of this stage were considered as initial conditions of a static structural analysis including the pressure defined by DBA. Finally, an equivalent static analysis was performed to analyze the seismic response considering the design basis spectra for the site. The different loads were combined and the abnormal/extreme environmental combination was the most unfavorable for the structure, defining the design.

  10. A Survey study on design procedure of Seismic Base Isolation ...

    African Journals Online (AJOL)

    Adding shear walls or braced frames can decrease the potential damage caused by earthquakes.We can isolate the structures from the ground using the Seismic Base Isolation Systems that is flexible approach to decrease the potential damage. In this research we present information on the design procedure of seismic ...

  11. Effects of applying three-dimensional seismic isolation system on the seismic design of FBR

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Kanazawa, Kenji; Matsuda, Akihiro

    1997-01-01

    In this study conceptional three-dimensional seismic isolation system for fast breeder reactor (FBR) is proposed. Effects of applying three-dimensional seismic isolation system on the seismic design for the FBR equipment are evaluated quantitatively. From the evaluation, it is concluded following effects are expected by applying the three-dimensional seismic isolation system to the FBR and the effects are evaluated quantitatively. (1) Reduction of membrane thickness of the reactor vessel (2) Suppression of uplift of fuels by reducing vertical seismic response of the core (3) Reduction of the supports for the piping system (4) Three-dimensional base isolation system for the whole reactor building is advantageous to the combined isolation system of horizontal base isolation for the reactor building and vertical isolation for the equipment. (author)

  12. Seismic analysis of the pile foundation of the reactor building of the NPP ANGRA 2

    International Nuclear Information System (INIS)

    Wolf, J.P.; Arx, G.A. von; Barros, F.C.P. de; Kakubo, M.

    1981-01-01

    A pile foundation subjected to dynamic loads interacts with the surrounding soil. Frequency-dependent stiffness and radiation damping must be properly taken into account in pile-soil-pile interaction. Assuming that the soil consists of horizontal layers of elastic material with hysteretic damping, the dynamic stiffness of a group of (even battered) piles can be determined, accounting rigorously for the cavities where the soil is subsequently replaced by the piles. By way of illustration, this substructure procedure, which works in the frequency domain, is applied to the final design of the pile foundation of the Reactor Building of Angra 2 in Brazil. Below the basemat, a strongly horizontally-layered compressive soil of 36 m thickness rests on bedrock. The reactor building is founded on 202 endbearing piles and 88 floating piles of 15 m length. Every pile is modelled. Along each pile, compatibility between the pile and the soil in all three directions is formulated in seven nodes. The basemat is assumed to be rigid. On the level of bedrock a broad-banded response spectrum specifies the excitation (outcropping). (orig./WL)

  13. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  14. Seismic isolation design guidelines for KALIMER(Revision A)

    International Nuclear Information System (INIS)

    Yoo, B; Koo, Gyeong Hoi; Lee, J. H.

    2000-04-01

    The main purpose of this report is to develop the seismic isolation design guideline for KALIMER(Korea Advanced LIquid MEtal Reactor). The proposed design rules(revision A) are only applicable to the seismic isolation design with using the high damping laminated rubber bearings. When using other seismic isolation devices and applying to 3-dimensional isolation, the proposed guidelines shall be modified and added with proper research data. The rules described in this report are based on the research results performed up to now but needed to be upgraded and verified with more detail research works for the future

  15. IDEF method for designing seismic information system in CTBT verification

    International Nuclear Information System (INIS)

    Zheng Xuefeng; Shen Junyi; Jin Ping; Zhang Huimin; Zheng Jiangling; Sun Peng

    2004-01-01

    Seismic information system is of great importance for improving the capability of CTBT verification. A large amount of money has been appropriated for the research in this field in the U.S. and some other countries in recent years. However, designing and developing a seismic information system involves various technologies about complex system design. This paper discusses the IDEF0 method to construct function models and the IDEF1x method to make information models systemically, as well as how they are used in designing seismic information system in CTBT verification. (authors)

  16. Preliminary seismic design of dynamically coupled structural systems

    International Nuclear Information System (INIS)

    Pal, N.; Dalcher, A.W.; Gluck, R.

    1977-01-01

    In this paper, the analysis criteria for coupling and decoupling, which are most commonly used in nuclear design practice, are briefly reviewed and a procedure outlined and demonstrated with examples. Next, a criterion judged to be practical for preliminary seismic design purposes is defined. Subsequently, a technique compatible with this criterion is suggested. A few examples are presented to test the proposed procedure for preliminary seismic design purposes. Limitations of the procedure are also discussed and finally, the more important conclusions are summarized

  17. Response and capacity evaluation of unit 1-2, Kozloduy NPP

    International Nuclear Information System (INIS)

    Kostov, M.; Stefanov, D.; Boncheva, H.

    1993-01-01

    Investigation described in this presentation was performed within the WANO Program for seismic safety assessment of Kozloduy NPP. The investigation is imposed by the necessity of seismic upgrading of the structures and equipment of the plant for the new design basic earthquake. Term of reference for this study was elaborated by experts of IAEA

  18. Sensitivity of seismic design parameters to input variables

    International Nuclear Information System (INIS)

    Wium, D.J.W.

    1987-01-01

    The probabilistic method introduced by Cornell (1968) has been used to a large extent for this purpose. Due to its probabilistic approach, this technique provides a sound basis for studying the influence of the dominant parameters in such a model. Although the Southern African region is not well known for its seismicity, a number of events in the recent past has focussed the attention on some seismically active areas where special attention may be needed in defining the correct design parameters. The relatively sparse historical seismic data has been used to develop a mathematical model which represents this region. This paper briefly discusses this model, and uses it as a basis for evaluating the influence of the uncertainty in each of the principal parameters, being the seismicity of the region, the attenuation of seismic waves after an event, and models that can be used to arrive at engineering design values. (orig./HP)

  19. The reevaluation of seismic safety of existing nuclear power plant

    International Nuclear Information System (INIS)

    Kitagawa, Hiroshi; Tominaga, Shohei; Kumagai, Chiyoshi; Koshiba, Koremutsu; Kono, Tomonori; Agawa, Kazuyoshi; Kuwata, Kenichiro

    2003-01-01

    We have carried out additional geological surveys in order to enrich our database on geological faults in the vicinity of Shimane Nuclear Power Plant (NPP). Prior to additional geological surveys, given the social importance of nuclear power plants, we hypothetically assumed that almost the whole length of an area covered by surveys would be an active fault that must be considered in seismic design, and tried to reevaluate the seismic safety of the NPP by applying an input earthquake ground motion larger than the level at the design stage. As a result, we have confirmed that seismic safety of the NPP can be maintained. This paper describes the method that we employed to reevaluate the seismic safety of Shimane NPP. (author)

  20. Seismic design of nuclear power plants - an assessment

    International Nuclear Information System (INIS)

    Howard, G.E.; Ibanez, P.; Smith, C.B.

    1976-01-01

    This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the 'nuclear power industry'. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation. (Auth.)

  1. Comparison of ex-USSR norms and current international practice in design of seismic resistant nuclear power plants

    International Nuclear Information System (INIS)

    Hauptenbuchner, B.; David, M.

    1995-01-01

    Seismic hazard has been estimated according to ex-USSR norms in the original designs of WWER type Nuclear Power Plants (NPP) in former Soviet Union as well as in all former east European countries. For some steps of the design the national standards has been also taken into account. The original ex-USSR norms and instructions has been several times changed and improved during the time. This contribution is dealing with the development of ex-USSR norms and regulations with the aim to recognise some most important differentiations in comparison with corresponding western or international ones from point of view of civil structures. The understanding of relations of these documents is very important for seismic qualification and upgrading of WWER-type, NPPs. The main Soviet/Russian Standards and Regulations related to the seismic design and qualification of NPP structures as SNiP II-A.12-69, VSN 15-78, SNiP II-7-81, PiNAE G-7-002-86, NTD SEV etc. have been taken into consideration and compared with western or international standards as IAEA 50-SG-S1, IAEA 50-SG-D15, KTA 2201.1-6, ASCE 4-86 etc. The numerical examples of structural seismic qualification has been elaborated according to different standards for better understanding and in order to determine the degree of safety referring to corresponding standards. The authors has tried also to take into account the way of application of ex-USSR norms. The comparison of different norms and regulations has been analysed and corresponding conclusions and recommendations have been derived. These conclusions and recommendations can be helpful by the seismic qualification and upgrading of WWER-type NPPs. (author)

  2. Update of the tectonic model for the Pannonian basin: a contribution to the seismic hazard reassessment of the Paks NPP (Hungary)

    Science.gov (United States)

    Horváth, Ferenc; Tóth, Tamás; Wórum, Géza; Koroknai, Balázs; Kádi, Zoltán; Kovács, Gábor; Balázs, Attila; Visnovitz, Ferenc

    2015-04-01

    The planned construction of two new units at the site of the Paks NPP requires a comprehensive site investigation including complete reassessment of the seismic hazard according to the Hungarian as well as international standards. Following the regulations of the Specific Safety Guide no. 9 (IAEA 2010), the approved Hungarian Geological Investigation Program (HGIP) includes integrated geological-geophysical studies at different scales. The regional study aims at to elaborate a new synthesis of all published data for the whole Pannonian basin. This task is nearly completed and the main outcomes have already been published (Horváth et al. 2015). The near regional study is in progress and addresses the construction of a new tectonic model for the circular area with 50 km radius around the NPP using a wealth of unpublished oil company seismic and borehole data. The site vicinity study has also been started with a core activity of 300 km² 3D seismic data acquisition, processing and interpretation assisted by a series of additional geophysical surveys, new drillings and geological mapping. This lecture will present a few important results of the near regional study, which sheds new light on the intricate tectonic evolution of the Mid-Hungarian Fault Zone (MHFZ), which is a strongly deformed belt between the Alcapa and Tisza-Dacia megatectonic units. The nuclear power plant is located at the margin of the Tisza unit near to the southern edge of the MHFZ. Reassessment of seismic hazard at the site of the NPP requires better understanding of the Miocene to Recent tectonic evolution of this region in the central part of the Pannonian basin. Early to Middle Miocene was a period of rifting with formation of 1 to 3 km deep half-grabens filled with terrestrial to marine deposits and large amount of rift-related volcanic material. Graben fill became strongly deformed as a consequence of juxtaposition of the two megatectonic units leading to strong compression and development of

  3. Seismic design and evaluation criteria based on target performance goals

    International Nuclear Information System (INIS)

    Murray, R.C.; Nelson, T.A.; Kennedy, R.P.; Short, S.A.

    1994-04-01

    The Department of Energy utilizes deterministic seismic design/evaluation criteria developed to achieve probabilistic performance goals. These seismic design and evaluation criteria are intended to apply equally to the design of new facilities and to the evaluation of existing facilities. In addition, the criteria are intended to cover design and evaluation of buildings, equipment, piping, and other structures. Four separate sets of seismic design/evaluation criteria have been presented each with a different performance goal. In all these criteria, earthquake loading is selected from seismic hazard curves on a probabilistic basis but seismic response evaluation methods and acceptable behavior limits are deterministic approaches with which design engineers are familiar. For analytical evaluations, conservatism has been introduced through the use of conservative inelastic demand-capacity ratios combined with ductile detailing requirements, through the use of minimum specified material strengths and conservative code capacity equations, and through the use of a seismic scale factor. For evaluation by testing or by experience data, conservatism has been introduced through the use of an increase scale factor which is applied to the prescribed design/evaluation input motion

  4. Seismic design criteria for special isotope separation plant structures

    International Nuclear Information System (INIS)

    Wrona, M.W.; Wuthrich, S.J.; Rose, D.L.; Starkey, J.

    1989-01-01

    This paper describes the seismic criteria for the design of the Special Isotope Separation (SIS) production plant. These criteria are derived from the applicable Department of Energy (DOE) orders, references and proposed standards. The SIS processing plant consistent of Load Center Building (LCB), Dye Pump Building (DPB), Laser Support Building (LSB) and Plutonium Processing Building (PPB). The facility-use category for each of the SIS building structures is identified and the applicable seismic design criteria and parameters are selected

  5. Cost reduction through improved seismic design

    International Nuclear Information System (INIS)

    Severud, L.K.

    1984-01-01

    During the past decade, many significnt seismic technology developments have been accomplished by the United States Department of Energy (USDOE) programs. Both base technology and major projects, such as the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR) plant, have contributed to seismic technology development and validation. Improvements have come in the areas of ground motion definitions, soil-structure interaction, and structural analysis methods and criteria for piping, equipment, components, reactor core, and vessels. Examples of some of these lessons learned and technology developments are provided. Then, the highest priority seismic technology needs, achievable through DOE actions and sponsorship are identified and discussed. Satisfaction of these needs are expected to make important contributions toward cost avoidances and reduced capital costs of future liquid metal nuclear plants. 23 references, 12 figures

  6. Optimum design for pipe-support allocation against seismic loading

    International Nuclear Information System (INIS)

    Hara, Fumio; Iwasaki, Akira

    1996-01-01

    This paper deals with the optimum design methodology of a piping system subjected to a seismic design loading to reduce its dynamic response by selecting the location of pipe supports and whereby reducing the number of pipe supports to be used. The author employs the Genetic Algorithm for obtaining a reasonably optimum solution of the pipe support location, support capacity and number of supports. The design condition specified by the support location, support capacity and the number of supports to be used is encored by an integer number string for each of the support allocation candidates and they prepare many strings for expressing various kinds of pipe-support allocation state. Corresponding to each string, the authors evaluate the seismic response of the piping system to the design seismic excitation and apply the Genetic Algorithm to select the next generation candidates of support allocation to improve the seismic design performance specified by a weighted linear combination of seismic response magnitude, support capacity and the number of supports needed. Continuing this selection process, they find a reasonably optimum solution to the seismic design problem. They examine the feasibility of this optimum design method by investigating the optimum solution for 5, 7 and 10 degree-of-freedom models of piping system, and find that this method can offer one a theoretically feasible solution to the problem. They will be, thus, liberated from the severe uncertainty of damping value when the pipe support guaranties the design capacity of damping. Finally, they discuss the usefulness of the Genetic Algorithm for the seismic design problem of piping systems and some sensitive points when it will be applied to actual design problems

  7. Impact of the measurement data on the CORD-2 nuclear design calculations of the NPP Krsko

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2004-01-01

    The CORD-2 package was developed at Jozef Stefan Institute and has been validated for the nuclear design calculations of PWR cores. It has been used for the independent verification of the NPP Krsko nuclear design for the last 6 cycles of operation. The accuracy of the package is very good fulfilling all criteria usually imposed on the design prediction of the reactor nuclear parameters. To obtain as robust package as possible and to eliminate potential systematic errors of the package, it was decided to rely on measured core power distributions. In core power measurements, which are performed each month of reactor operation, are used to obtain fuel assemblies burnup histories. Consequently, burnup distributions obtained from the power measurements of all previous cycles are taken as a starting point at the beginning of the considered cycle. Since a lot of experience has been gained with the package, it was decided to evaluate the impact of measurement data on the accuracy of the calculations. Burnup calculations of all 19 cycles of the NPP Krsko are repeated, building simultaneously the calculated library of burnup histories for all fuel assemblies. The basic reactor parameters such as HZP critical boron concentration, isothermal temperature coefficient, control rod worth and cycle length are compared to the results obtained with CORD-2 standard sequence of calculation and direct measurements.(author)

  8. The 18 basic requirement of quality assurance for American design NPP

    International Nuclear Information System (INIS)

    Baliza, Ana Rosa

    2013-01-01

    On April 17th, 1969, the Atomic Energy Commission (AEC) published in the U.S. Federal Register (FR), Volume 34, Number 73, a proposed amendment to 10CFR50 to insert Appendix B - 'Quality assurance criteria for Nuclear Power Plant'. This Appendix was officially approved on June 27th, 1970 and published in the FR, volume 35, number 125. Appendix B is the Quality Assurance document for U.S. nuclear facilities. This document establishes eighteen basic requirements (BR) to design, construction, manufacture and operation of structures, systems and components (SSC) related to safety. The 18 BR describe 'what' shall be done, but not 'how' to do. In order to standardize the actions of nuclear facilities during 10 CFR 50 App B implementation, the industry has developed some documents, the main ones are: ASME NQA-1 (Quality Assurance Requirements for Nuclear Facility Applications) and the series ANSI N 45.2 (Quality Assurance Program Requirements for Nuclear Facilities). Both documents are approved by the NRC (Nuclear Regulatory Commission). The NRC is the licensing body of U.S. nuclear facilities. In Brazil, the licensing body is CNEN (Comissao Nacional de Energia Nuclear). This paper describes the 18 BR for American Designed Nuclear Power Plant (NPP), applicable to Angra-1 NPP. (author)

  9. Development of tools to manage the operational monitoring and pre-design of the NPP-LV cycle

    International Nuclear Information System (INIS)

    Perusquia, R.; Arredondo S, C.; Hernandez M, J. L.; Montes T, J. L.; Castillo M, A.; Ortiz S, J. J.

    2015-09-01

    This paper presents the development of tools to facilitate the management so much, the operational monitoring of boiling water reactors (BWR) of the nuclear power plant of Laguna Verde (NPP-LV) through independent codes, and how to carry out the static calculations corresponding to process of optimized pre-design of the reference cycle next to current cycle. The progress and preliminary results obtained with the program SACal, developed at Instituto Nacional de Investigaciones Nucleares (ININ), central tool to achieve provide a management platform of the operational monitoring and pre-design of NPP-LV cycle are also described. The reached preliminary advances directed to get an Analysis center and automated design of fuel assembly cells are also presented, which together with centers or similar modules related with the fuel reloads form the key part to meet the targets set for the realization of a Management Platform of Nuclear Fuel of the NPP-LV. (Author)

  10. Seismic design method of free standing rack

    International Nuclear Information System (INIS)

    Taniguchi, Katsuhiko; Okuno, Daisaku; Iwasaki, Akihisa; Nekomoto, Yoshitsugu; Matsuoka, Toshihiro

    2013-01-01

    For high earthquake resistance and ease of installation, free standing racks which are not anchored to the pool floor or walls has been adopted in many countries. Under the earthquake, the response of the free standing rack is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning) and impacts between the fuel assemblies and the fuel cell walls, rack-to-rack, and the pit floor and rack pedestals. We carried out seismic experiments on the full-scale rack model in water and dry conditions to obtain the fundamental data about free standing rack (sliding, rocking and turning motions). We have developed the nonlinear dynamic analysis method to predict seismic response for the free standing rack utilizing the full-scale test result and verified the analysis evaluation method of the rack by comparison of test result. (author)

  11. Seismic analysis response factors and design margins of piping systems

    International Nuclear Information System (INIS)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01

    The objective of the simplified methods project of the Seismic Safety Margins Research Program is to develop a simplified seismic risk methodology for general use. The goal is to reduce seismic PRA costs to roughly 60 man-months over a 6 to 8 month period, without compromising the quality of the product. To achieve the goal, it is necessary to simplify the calculational procedure of the seismic response. The response factor approach serves this purpose. The response factor relates the median level response to the design data. Through a literature survey, we identified the various seismic analysis methods adopted in the U.S. nuclear industry for the piping system. A series of seismic response calculations was performed. The response factors and their variabilities for each method of analysis were computed. A sensitivity study of the effect of piping damping, in-structure response spectra envelop method, and analysis method was conducted. In addition, design margins, which relate the best-estimate response to the design data, are also presented

  12. Design of Evaporator for Liquid Radioactive Waste Treatment-NPP 1000 MW, PWR

    International Nuclear Information System (INIS)

    Mulyono Daryoko

    2007-01-01

    The evaporator for liquid radioactive waste treatment of 1000 MW NPP-PWR has been designed. The basic calculate of this design was capacity 7000 l/hr, which 5 mg/l solid content. The system used was superheated steam 3.4 atmosphere, 281°F. The data required from design of evaporator are evaporator part (heat exchanger): diameter of shell is 33 inch (82.50 cm), nominal diameter tube is 1.5 inch (3.75 cm), number of tube is 215, tube arrangement triangular pitch is 1 inch pitch, height 600 cm; Mist separator: diameter is 200 inch (500 cm), height 600 inch (1500 cm); Condenser: diameter of shell is 33 inch (82.50 cm), nominal diameter tube is 1.5 inch (3.75 cm), number of tube is 215, tube arrangement triangular pitch is 1 inch pitch, height 600 cm. (author)

  13. Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

    International Nuclear Information System (INIS)

    Terasvirta, R.

    2001-01-01

    This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

  14. Rupture of DN 500 - design basic accident at units 3 and 4 of Kozloduy NPP

    International Nuclear Information System (INIS)

    Uruchev, V.; Vassilev, P.; Ivanova, A.; Sartmadjiev, A.

    2005-01-01

    The original design of Kozloduy NPP Units 3 and 4 assumes as Design Basis Accident (DBA) the rupture of DN 32 mm primary pipeline, while an initial event of double-sided guillotine break of primary pipeline with maximal diameter is not considered. In the course of units modernization it have been demonstrated once and again that both the emergency core cooling systems and the localization systems can cope with larger and larger primary circuit leaks. After the installation of a Jet-Vortex Condenser (JVC) at Units 3 and 4 it was substantiated that, the integrity of the hermetic rooms is ensured even in case of double-sided guillotine break of a primary circuit pipeline with maximal diameter (DEGB). The technical justification of the jet-vortex condenser, elaborated by VNIAEC, contains calculations determining both the source term and the doses obtained outside the NPP site after LOCA DN 500. LOCA DN 500 is considered in these analyses as a beyond design basis accident and it is so included in the SAR and approved by the Nuclear Regulatory Agency (NRA). The thermo-hydraulic calculations performed later on show that the emergency core cooling systems can cope with this initial event at conservative assumptions. In order to classify this initiating event as a design basis accident it is necessary to demonstrate that the core cooling criteria are fulfilled and the internal and external doses outside the NPP site are within the permissible limits fixed for design basis accident by the Bulgarian regulatory body (NRA), when using conservative assumptions. For this purpose two consecutive studies were performed - evaluation of the DEGB probability and categorization of the initial event according to the contemporary regulations acting in Republic of Bulgaria. The presented report summarizes the results of the performed conservative analyses of double-sided guillotine break accident of main circulation line taking into account the probability of rupture of large diameter

  15. Seismic hazard maps for earthquake-resistant construction designs

    International Nuclear Information System (INIS)

    Ohkawa, Izuru

    2004-01-01

    Based on the idea that seismic phenomena in Japan varying in different localities are to be reflected in designing specific nuclear facilities in specific site, the present research program started to make seismic hazard maps representing geographical distribution of seismic load factors. First, recent research data on historical earthquakes and materials on active faults in Japan have been documented. Differences in character due to different localities are expressed by dynamic load in consideration of specific building properties. Next, hazard evaluation corresponding to seismic-resistance factor is given as response index (spectrum) of an adequately selected building, for example a nuclear power station, with the help of investigation results of statistical analysis. (S. Ohno)

  16. Refer to AP1000 for discussing the betterment of seismic design of internal nuclear power plant

    International Nuclear Information System (INIS)

    Gong Zhenbang; Zhang Renyan

    2014-01-01

    As a reference technique of AP1000, This paper discussed the betterment of seismic design of nuclear power plant in three ways. (1) Establish design criteria and guidelines for protection from seismic interaction; (2) Nuclear power plant seismic design of eliminating or weaken operation-basis earthquake; (3) Develop the seismic margin analysis (SMA) of the nuclear power plant. These three aspect are frontier technology in internal seismic design of internal nuclear power plant, and also these three technology are related intimately. (authors)

  17. Optimal design of water supply networks for enhancing seismic reliability

    International Nuclear Information System (INIS)

    Yoo, Do Guen; Kang, Doosun; Kim, Joong Hoon

    2016-01-01

    The goal of the present study is to construct a reliability evaluation model of a water supply system taking seismic hazards and present techniques to enhance hydraulic reliability of the design into consideration. To maximize seismic reliability with limited budgets, an optimal design model is developed using an optimization technique called harmony search (HS). The model is applied to actual water supply systems to determine pipe diameters that can maximize seismic reliability. The reliabilities between the optimal design and existing designs were compared and analyzed. The optimal design would both enhance reliability by approximately 8.9% and have a construction cost of approximately 1.3% less than current pipe construction cost. In addition, the reinforcement of the durability of individual pipes without considering the system produced ineffective results in terms of both cost and reliability. Therefore, to increase the supply ability of the entire system, optimized pipe diameter combinations should be derived. Systems in which normal status hydraulic stability and abnormal status available demand could be maximally secured if configured through the optimal design. - Highlights: • We construct a seismic reliability evaluation model of water supply system. • We present technique to enhance hydraulic reliability in the aspect of design. • Harmony search algorithm is applied in optimal designs process. • The effects of the proposed optimal design are improved reliability about by 9%. • Optimized pipe diameter combinations should be derived indispensably.

  18. Study on design method for seismically isolated FBR plants

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

    1998-01-01

    CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)

  19. Optimization study and preliminary design for Latina NPP early core retrieval and reactor dismantling

    International Nuclear Information System (INIS)

    Macci, E.; Zirpolo, S.; Imparato, A.; Cacace, A.; Parry, D.; Walkden, P.

    2002-01-01

    In June 2000, an agreement was established between Sogin and BNFL to enable the two companies to co-operate, using their specific experiences in the decommissioning field, for the benefit of projects in Italy, the United Kingdom and for third markets. A decommissioning strategy for the Latina NPP was initially developed in a Phase 1 Study which produced a conceptual design for the decommissioning of the reactor. This study was completed in June 2000. Since then, a second study has been completed, which has further developed the strategy and produced preliminary designs for the early dismantling of the core and reactor building at Latina. The engineering and safety data were produced in order to support Sogin in the preparation of a safety case for plant decommissioning. This safety case was submitted to the Italian Regulator, ANPA, in February 2002. (author)

  20. Seismic strengthening of nuclear power plants V1-V2 structures in Slovak Republic

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

    The structural upgrading of main buildings of Bohunice NPP units V1 and V2 is described in this presentation. Design criteria for structural upgrading are included. Since the seismic upgrading of the existing NPP is usually very complicated and expensive task, designer is obliged to find the optimal solution between the economics and reliability of the upgrading. The assistance of IAEA missions during the process of Bohunice seismic upgrading is considered very fruitful

  1. Update of bridge design standards in Alabama for AASHTO LRFD seismic design requirements.

    Science.gov (United States)

    2013-11-01

    The Alabama Department of Transportation (ALDOT) has been required to update their bridge design to the LRFD Bridge Design Specifications. This transition has resulted in changes to the seismic design standards of bridges in the state. Multiple bridg...

  2. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

    PAKS NPP was built according to Soviet construction standards which took into account meteorological aspects but no documents for other external hazards were available. Main activities concerning earthquakes cover reevaluation of the plant site, seismic safety technological concept, improving the seismic resistance, installation of seismic monitoring and protection system, and seismic PSA

  3. The-Abstraction-Hierarchy-based Mobile PC Display Design for NPP Maintenance

    International Nuclear Information System (INIS)

    Kim, In; Kim, Bo Gyung; Seong, Poong Hyun; Ha, Jun Su

    2010-01-01

    Recently, the importance of effective maintenance in nuclear power plants (NPPs) has been emphasized and research into effective maintenance by adopting mobile maintenance aids (MMAs) have been attempted. For improved and effective use of an MMA display design method based on abstraction hierarchy (AH) is proposed and its design considerations are discussed in this study. Six levels of abstraction hierarchy are proposed in this paper to classify the maintenance information. By classifying and organizing maintenance information using AH, maintenance information can be used effectively by users with either high or low levels of expertise. When information classification has been finished, the information for MMA design is selected and designed. With the considerations of MMA design analysis and guidelines, AH-based MMA is designed for the maintenance tasks. An experiment is conducted using the AH-based MMA in order to estimate the effectiveness of the proposed method for the maintenance tasks and to identify design considerations to enhance the proposed MMAs. The result indicated that an AH-based manual was more effective than a conventional manual in terms of task completion time and number of errors. The workload for the AH-based manual was estimated less than the conventional manual for subjects with low level of expertise. As the level of expertise increases, subjects tended to follow more abstract information while the number of navigations decreased. It is believed that when mobile devices become pervasive in NPP maintenance fields, AH-model applied MMAs can be used as an effective maintenance supporting tool

  4. Seismic assessment and upgrading of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Tamas, K.

    2001-01-01

    A comprehensive programme for seismic assessment and upgrading is currently in progress at Hungary's Paks NPP. The re-evaluation of the site seismic hazard had been already completed. The technology of safe shut down and heat removal is established and the systems and structures relevant for seismic safety are identified. A seismic instrumentation is installed. The pre-earthquake preparedness and post-earthquake actions are elaborated. The methods for seismic capacity assessment are selected. The seismic capacity evaluation and the design of upgrading measures are currently in progress. The easy to perform upgrading covering the most urgent measures had been already performed. (author)

  5. Environmental qualification design for NPP refurbishment to comply with revised licensing requirements

    International Nuclear Information System (INIS)

    MacBeth, M. J.; Hemmings, R. L.

    2002-01-01

    Recent Canadian Nuclear Regulatory decisions have imposed Environmental Qualification (EQ) requirements for twenty-four Reactor Building (RB) airlocks at the four-unit Pickering Nuclear Generating Station-B (PNGS-B) facility. This paper describes the EQ modification design work completed by CANATOM-NPM for the problematic aspects for such projects. The airlocks allow RB access while providing a containment boundary and are designed to prevent a potential breach of containment for all analysed station conditions. Each PNGS-B unit has three large equipment airlocks and three smaller personnel airlocks. The airlocks must function under postulated worst-case design basis accident(DBA) conditions for assigned mission durations. The design must ensure that accident conditions cannot spuriously initiate an un-requested door opening. CANATOM-NPM reviewed site data to specify the necessary EQ modifications required to satisfy licensing requirements while providing a correct and complete as-found record of the existing airlock installation. The design team assessed the installed airlocks configuration against environmental qualification requirements to finalize the list of necessary modifications. A comprehensive, cross-discipline review of proposed design changes was completed to identify any further changes required to satisfy the final EQ licensing goal. The design team also conducted a design review of the EQ modification installation strategy to integrate the design deliverables with the installation team requirements while attempting to minimize necessary outage time for EQ modification installations. This project was completed on schedule and within the cost limitations required by the client with comprehensive, high quality final design packages. Overall improvements were realized for OPG system drawings and the electronic documentation of design data. The EQ modifications designed by CANATOM-NPM will ensure the continued operation of the PNGS-B NPP past December 31

  6. Peer review for conceptual design of the new safe confinement for the Chernobyl NPP shelter object

    International Nuclear Information System (INIS)

    Kupny, Valentin; Shestopalov, Vyacheslav; Sobotovich, Emlen; Tokarevsky, Vladimir; Veryuzhsky, Yuri; Abdulakhatov, Murat

    2005-01-01

    The results of peer review for Conceptual Design of the New Safe Confinement (NSC) for Chernobyl NPP Shelter Object in the Arch option are presented. NSC consists of: 1) main building, including steel arch structure of tubular trusses, covered with thin-sheet metal (its bay in the direction north-south is equal to 257.44 m, height - 108.39 m, length - 150 m), foundations, western and eastern front walls; 2) technological (process) building, including sites for decontamination, fragmentation and packaging, sanitary locks, workshops and other technological premises; 3) auxiliary systems and structures. The following questions are considered: evolution of the requirements to the new Shelter-2, compliance of functional and engineering solutions; compliance with normative documents, standards and laws. The Arch design has no advantages compared with other known options for SO transformation into an ecologically safe system: by its process capabilities, it yields to the Dock-Caisson design; by cost of construction and operational expenses, it yields to the 'Monolith design; by dose expenses for construction and strength parameters it yields to the 'Rainbow design. (author)

  7. Results of a benchmark study for the seismic analysis and testing of WWER type NPPs: Overview and general comparison for Paks NPP

    International Nuclear Information System (INIS)

    Guerpinar, A.; Zola, M.

    2001-01-01

    Within the framework of the IAEA coordinated 'Benchmark Study for the seismic analysis and testing of WWER-type NPPs', in-situ dynamic structural testing activities have been performed at the Paks Nuclear Power Plant in Hungary. The specific objective of the investigation was to obtain experimental data on the actual dynamic structural behaviour of the plant's major constructions and equipment under normal operating conditions, for enabling a valid seismic safety review to be made. This paper refers on the comparison of the results obtained from the experimental activities performed by ISMES with those coming from analytical studies performed for the Coordinated Research Programme (CRP) by Siemens (Germany), EQE (Bulgaria), Central Laboratory (Bulgaria), M. David Consulting (Czech Republic), IVO (Finland). This paper gives a synthetic description of the conducted experiments and presents some results, regarding in particular the free-field excitations produced during the earthquake-simulation experiments and an experiment of the dynamic soil-structure interaction global effects at the base of the reactor containment structure. The specific objective of the experimental investigation was to obtain valid data on the dynamic behaviour of the plant's major constructions, under normal operating conditions, to support the analytical assessment of their actual seismic safety. The full-scale dynamic structural testing activities have been performed in December 1994 at the Paks (H) Nuclear Power Plant. The Paks NPP site has been subjected to low level earthquake-like ground shaking, through appropriately devised underground explosions, and the dynamic response of the plant's 1st reactor unit important structures was appropriately measured and digitally recorded, with the whole nuclear power plant under normal operating conditions. In-situ free field response was measured concurrently and, moreover, site-specific geophysical and seismological data were simultaneously

  8. Evaluation of seismic criteria used in design of INEL facilities

    International Nuclear Information System (INIS)

    Young, G.A.

    1977-01-01

    This report provides the results of an independent evaluation of seismic studies that were made to establish the seismic acceleration levels and the response spectra used in the design of vital facilities at Idaho National Engineering Laboratory. A comparison of the procedures used to define the seismic acceleration values and response spectra at INEL with the requirements of the Nuclear Regulatory Commission showed that additional geologic studies would probably be required in order to fulfill NRC regulations. Recommendations are made on justifiable changes in the acceleration values and response spectra used at INEL. The geologic, geophysical, and seismological studies needed to provide a better understanding of the tectonic processes in the Snake River plains and the surrounding region are identified. Both potential and historical acceleration values are evaluated on a probability basis to permit a risk assessment approach to the design of new facilities and facility modifications. Studies conducted to develop seismic criteria for the design of the Loss of Fluid Test reactor and the New Waste Calcining Facility were selected as typical examples of criteria development previously used in the design of INEL facilities

  9. Seismic design considerations of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2001-10-01

    An Advisory Group Meeting (AGM) on Seismic Technologies of Nuclear Fuel Cycle Facilities was convened in Vienna from 12 to 14 November 1997. The main objective of the meeting was the investigation of the present status of seismic technologies in nuclear fuel cycle facilities in Member States as a starting point for understanding of the most important directions and trends of national initiatives, including research and development, in the area of seismic safety. The AGM gave priority to the establishment of a consistent programme for seismic assessment of nuclear fuel cycle facilities worldwide. A consultants meeting subsequently met in Vienna from 16 to 19 March 1999. At this meeting the necessity of a dedicated programme was further supported and a technical background to the initiative was provided. This publication provides recommendations both for the seismic design of new plants and for re-evaluation projects of nuclear fuel cycle facilities. After a short introduction of the general IAEA approach, some key contributions from Member State participants are presented. Each of them was indexed separately

  10. Site-specific issues related to structural/seismic design of an underground independent spent fuel storage installation (ISFSI)

    International Nuclear Information System (INIS)

    Tripathi, B.P.

    2005-01-01

    Utilities owning and operating commercial nuclear power plants (NPP) in USA may choose to build an underground Independent Spent Fuel Storage Installation (ISFSI) to store the spent nuclear fuels. The regulatory requirements and other guidance are based on 10 CFR Part 72, Regulatory Guide RG 3.73, Standard Review Plans NUREG-1536 and NUREG-1567, and Interim staff Guidance (ISG) documents as applicable. Structures, Systems, and Components (SSCs) classified as important to safety are designed to withstand the effects of site-specific environmental conditions and natural phenomena such as earthquake, tornado, flood, etc. An underground ISFSI for storage of spent nuclear fuel, presents some unique analysis and design challenges. This paper will briefly address some of these challenges and discuss site-specific loads, including seismic for the ISFSI design. (authors)

  11. The environmental constraint needs for design improvements to the Saligny I/LLW-repository near Cernavoda NPP

    International Nuclear Information System (INIS)

    Barariu, Gheorghe

    2007-01-01

    The paper presents the new perspectives on the development of the L/ILW Final Repository Project which will be built near Cernavoda NPP. The Repository is designed to satisfy the main performance objectives in accordance to IAEA recommendation. Starting in October 1996, Romania became a country with an operating nuclear power plant. Reactor 2 reached the criticality on May 6, 2007 and it will be put in commercial operation in September 2007. The Ministry of Economy and Finance has decided to proceed with the commissioning of Units 3 and 4 of Cernavoda NPP till 2014. The Strategy for radioactive waste management was elaborated by National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for definitive disposal and the coordination of nuclear spent fuel and radioactive waste management (Order 844/2004) with attributions established by Governmental Decision (GO) 31/2006. The Strategy specifies the commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP in 2014. When designing the L/IL Radwaste Repository, the following prerequisites have been taken into account: 1) Cernavoda NPP will be equipped with 4 Candu 6 units. 2) National Legislation in radwaste management will be reviewed and/or completed to harmonize with UE standards 3) The selected site is now in process of confirmation after a comprehensive set of interdisciplinary investigations. (author)

  12. Design of an integrated operator support system for advanced NPP MCRs. Issues and perspectives

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Seong Poong-Hyun

    2010-01-01

    Recently, human error has been highlighted as one of the main causes of accidents in nuclear power plants (NPPs). In order to prevent human errors during the main control room (MCR) operations, which are highly complex and mentally taxing activities, improved interfaces and operator support systems have been developed for advanced MCRs. Although operator support systems have the capability to improve the safety and reliability of an NPP, inappropriate designs can have adverse effects on the system safety. Designs based on systematic development frames and validation/verification of the systems are pivotal strategies to circumvent the negative effects of operator support systems. In this paper, an integrated operator support system designed to aid the cognitive activities of operators as well as theoretical and experimental evaluation methods of operator support systems are reviewed. From this review, it was concluded that not only issues about systems (e.g., the accuracy of the system outputs), but also issues about human operators who use the systems (for instance, information quality, the operator's trust and dependency on support systems) should be considered in the design of efficient operator support systems. (author)

  13. Historical development of the seismic requirements for construction of nuclear power plants in the U.S. and worldwide and their current impact on cost and safety

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    2003-01-01

    The following topics are described and discussed: Historical development of NPP seismic design requirements: Peak ground acceleration; Response spectra and damping; Floor or amplified response spectra; Effective high frequency response spectra; Seismic modeling procedures; Impact on cost (site preparation and foundations; site seismic response and generation of site dependent spectra). Potential use of indirect earthquake experience data in design and construction of NPP. Seismic contribution to safety. The following facts are summarized in two Appendices: Seismic intensity scales, and GRS safety codes and guides. (P.A.)

  14. Design and development of indigenous seismic switch for nuclear reactors

    International Nuclear Information System (INIS)

    Varghese, Shiju; Shah, Jay; Limaye, P.K.; Soni, N.L; Patel, R.J.

    2016-01-01

    After Fukushima incident it has become a regulatory requirement to have automatic reactor trip on detection of earthquake beyond OBE level. Seismic Switches that meets the technical specifications required for nuclear reactor use were not available in the market. Hence, on Nuclear Power Corporation of India Ltd (NPCIL's) request, Refuelling Technology Division, BARC has developed Seismic Switches (electronic earthquake detectors) required for this application. Functionality of the system was successfully tested using a Shake Table. Two different designs of seismic switches have been developed. One is a microcontroller based system (digital) and the other is fully analogue electronics (analog) based. These switches are designed to meet the technical requirements of Class IA systems of nuclear reactors. It is also designed to meet other qualification tests such as EMI/EMC, climatic, vibration, and reliability requirements. In addition to nuclear industry seismic switches are having potential use in oil and gas, power plants, buildings and other industrial installations. These technologies are currently available for technology transfer and details are published in BARC website. This paper describes the requirements, principle of operation and features and testing of the developed systems. (author)

  15. Seismic design for Monju FBR power plant

    International Nuclear Information System (INIS)

    1982-01-01

    This technical report introduces the basic concept on the aseismatic design of the FBR ''Monju'' power station, of which the construction in Tsuruga is planned by the Power Reactor and Nuclear Fuel Development Corp. The safety design of Monju has been performed according to ''The concept of evaluating the safety of fast breeder reactors'', and the thought concerning the aseismatic design also is written in it. According to it, ''The guide for the examination of aseismatic design regarding power reactor facilities'' should be referred to, and the classification according to the importance in aseismatic design must be made, taking the features in the design of liquid metal-cooled FBRs fully in consideration. In the aseismatic design of Monju performed according to these basic concept, the following two points were examined. In the aseismatic design of the equipment and piping, the difference of construction from LWRs such as low pressure, thin walled and high temperature construction is taken in consideration. The classification according to the aseismatic importance of the system and equipment is made on the basis of the features in the design of Monju. The classification according to aseismatic importance, the method of calculating earthquake power, the combination of loads and the allowable limit, and the aseismatic construction of the main facilities are reported. (Kako, I.)

  16. An Abstraction Hierarchy based mobile PC display design in NPP maintenance considering the level of expertise

    International Nuclear Information System (INIS)

    Yim, Ho Bin; Kim, In; Seong, Poong Hyun

    2011-01-01

    Research highlights: → Six levels of Abstraction Hierarchy based information for maintenance were proposed. → Errors and workload with AH based information display were reduced for LL subjects. → Design concerns discovered can be applied to practical use of mobile maintenance aids. - Abstract: Recently, the importance of effective maintenance in nuclear power plants (NPPs) has been emphasized and research into effective maintenance by adopting mobile maintenance aids (MMAs) have been attempted. For improved and effective use of an MMA display design method based on the hierarchy is proposed and its design considerations are discussed in this study. Six levels of hierarchy are proposed in this paper to classify the maintenance information. By classifying and organizing maintenance information using the hierarchy, maintenance information can be used effectively by users with either high or low levels of expertise. When information classification has been finished, the information for MMA design is selected and designed. With the considerations of MMA design analysis and guidelines, a hierarchy-based MMA is designed for the maintenance tasks. An experiment is conducted using the hierarchy-based MMA in order to estimate the effectiveness of the proposed method for the maintenance tasks and to identify design considerations to enhance the proposed MMAs. The result indicated that a hierarchy-based manual was more effective than a conventional manual in terms of task completion time and number of errors. The workload for the hierarchy-based manual was estimated less than the conventional manual for subjects with low level of expertise. As the level of expertise increases, subjects tended to follow more abstract information while the number of navigations decreased. It is believed that when mobile devices become pervasive in NPP maintenance fields, the hierarchy model applied MMAs can be used as an effective maintenance supporting tool.

  17. Appraisal of the implementation status of appendix general design criteria, of the 10CFR50 in the design of the Juragua NPP

    International Nuclear Information System (INIS)

    Rodriguez Aleman, Carlos; Mitjans Sanchez, Guillermo

    1996-01-01

    The work intends to reflect in a general manner the state of accomplishment of the general design requirements for NPP in the US and other parts of the world contained in 10CFR50 appendix A in the design solutions adopted for the Juragua plant with reactors VVER-440 V-318

  18. Seismic analyses of Paks RB. Progress report 1993-1994

    Energy Technology Data Exchange (ETDEWEB)

    David, M [David Consulting, Engineering and Design Office (Czech Republic)

    1995-07-01

    The dynamic analysis presented in this report refers to the seismic analysis of the main building of Paks NPP. The following tasks which have been completed are described: design of 3-dimensional model of the main building; calculation of frequencies and modes of free vibrations; determination of modal masses for all modes of vibrations; floor response spectra as response to seismic excitation assumed for the Paks site; relative response of seismic acceleration at the top of the condensing tower.

  19. Seismic Data Gathering and Validation

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Three recent earthquakes in the last seven years have exceeded their design basis earthquake values (so it is implied that damage to SSC’s should have occurred). These seismic events were recorded at North Anna (August 2011, detailed information provided in [Virginia Electric and Power Company Memo]), Fukushima Daichii and Daini (March 2011 [TEPCO 1]), and Kaswazaki-Kariwa (2007, [TEPCO 2]). However, seismic walk downs at some of these plants indicate that very little damage occurred to safety class systems and components due to the seismic motion. This report presents seismic data gathered for two of the three events mentioned above and recommends a path for using that data for two purposes. One purpose is to determine what margins exist in current industry standard seismic soil-structure interaction (SSI) tools. The second purpose is the use the data to validated seismic site response tools and SSI tools. The gathered data represents free field soil and in-structure acceleration time histories data. Gathered data also includes elastic and dynamic soil properties and structural drawings. Gathering data and comparing with existing models has potential to identify areas of uncertainty that should be removed from current seismic analysis and SPRA approaches. Removing uncertainty (to the extent possible) from SPRA’s will allow NPP owners to make decisions on where to reduce risk. Once a realistic understanding of seismic response is established for a nuclear power plant (NPP) then decisions on needed protective measures, such as SI, can be made.

  20. Seismic considerations in the design of atomic power plants

    International Nuclear Information System (INIS)

    Arya, A.S.; Chandrasekaran, A.R.; Thakkar, S.K.

    1975-01-01

    A seismic design is one of the most important factors for the safety of nuclear power plants constructed in seismic areas. The various considerations in the design of atomic power plant structures and components to achieve high degree (near absolute) of safety during future probable earthquakes is described as follows: (a) determination of design earthquake parameters for SSE and OBE (b) fixing time history accelerograms and acceleration response spectra (c) mathematical modelling of the reactor building considering soil-structure interaction (d) deciding allowable stresses, damping factors and serviceability limits like drift, displacements and crack widths (e) tests for determining stiffness and damping characteristics of components in-situ before commissioning of plant. The main questions that arise under various items requiring further research investigations or development work are pointed out for discussion. (author)

  1. Rigid-plastic seismic design of reinforced concrete structures

    DEFF Research Database (Denmark)

    Costa, Joao Domingues; Bento, R.; Levtchitch, V.

    2007-01-01

    structural strength with respect to a pre-defined performance parameter using a rigid-plastic response spectrum, which is characteristic of the ground motion alone. The maximum strength demand at any point is solely dependent on the intensity of the ground motion, which facilitates the task of distributing......In this paper a new seismic design procedure for Reinforced Concrete (R/C) structures is proposed-the Rigid-Plastic Seismic Design (RPSD) method. This is a design procedure based on Non-Linear Time-History Analysis (NLTHA) for systems expected to perform in the non-linear range during a lifetime...... earthquake event. The theoretical background is the Theory of Plasticity (Rigid-Plastic Structures). Firstly, a collapse mechanism is chosen and the corresponding stress field is made safe outside the regions where plastic behaviour takes place. It is shown that this allows the determination of the required...

  2. A study on mobile PC display design in NPP maintenance considering the level of expertise

    International Nuclear Information System (INIS)

    Kim, In

    2010-02-01

    Recently, the importance of effective maintenance in nuclear power plants (NPPs) has been emphasized and research into effective maintenance by adopting mobile maintenance aids (MMAs) have been attempted. MMAs are currently used during operation and maintenance in NPPs, but no method that considers the limitations of mobile devices has been proposed. For improved and effective use, an MMA display design method based on abstraction hierarchy (AH) is proposed and its design considerations are discussed in this study. Six levels of abstraction hierarchy are proposed in this paper to classify the maintenance information. By classifying and organizing maintenance information using AH, maintenance information can be used effectively by users either high or low levels of expertise. When information classification has been finished, the information requirements and relationships for MMA design is extracted from the analysis of the AH result. Representative human-machine interface (HMI) guidelines issued by the US NRC, which are generally used in NPPs, are applied. With the considerations of MMA design analysis and practical guidelines, AH-based MMA is designed for maintenance tasks. An experiment is conducted using the AH-based MMA in order to estimate the effectiveness of the proposed method for maintenance tasks and to identify design considerations to enhance the proposed MMAs. The result indicated that an AH-based manual was more effective than a conventional manual in terms of task completion time and number of errors. The workload for the AH-based manual was estimated to be less than the conventional manual for low levels of expertise. However, a conventional manual was more comprehensive than the proposed manual and the steps contained in the maintenance manual were easier to remember. As the level of expertise increases, subjects tended to follow more abstract information while the number of navigations decreased. It is believed that when mobile devices become

  3. Procedure for seismic evaluation and design of small bore piping

    International Nuclear Information System (INIS)

    Bilanin, W.; Sills, S.

    1991-01-01

    Simplified methods for the seismic design of small bore piping in nuclear power plants have teen used for many years. Various number of designers have developed unique methods to treat the large number of class 2 and 3 small bore piping systems. This practice has led to a proliferation of methods which are not standardized in the industry. These methods are generally based on enveloping the results of rigorous dynamic or conservative static analysis and result in an excessive number of supports and unrealistically high support loadings. Experience and test data have become available which warranted taking another look at the present methods for analysis of small bore piping. A recently completed Electric Power Research Institute and NCIG (a utility group) activity developed a new procedure for the seismic design and evaluation of small bore piping which provides significant safety and cost benefits. The procedure streamlines the approach to inertial stresses, which is the main feature that achieves the new benefits. Criteria in the procedure for seismic anchor movement and support design are based analysis and focus the designer on credible failure mechanisms. A walkdown of the as-constructed piping system to identify and eliminate undesirable piping features such as adverse spatial interaction is required

  4. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

  5. Seismic evaluation of non-seismically designed existing Magnox nuclear power plants

    International Nuclear Information System (INIS)

    Kunar, R.R.

    1984-01-01

    The philosophy and method adopted for the seismic assessment of three existing Magnox nuclear stations in the United Kingdom are presented in this paper. The plants were not seismically designed. The particular procedures that were applied were tailored to suit the difficulties of lack of data which is somewhat inevitable for plants designed and built about 25 to 30 years ago. Special procedures included on-site testing with a portable shake table, low vibration testing using a structural dynamics analyser, and on-site inspections. The low vibration testing was most invaluable in detecting differences between 'as-built' conditions and the engineering drawings. From the point of view of economics, this was more effective than conducting full structural surveys to determine the as-built conditions. The testing results also provided confidence in the answers from numerical models. The philosophy adopted for the Magnox reactors in the seismic assessment was to determine what peak ground accelerations the sites can sustain and then evaluate the chances of exceeding the ground accelerations over the remaining lifetime of the plants. The peak ground acceleration for each site was determined on the basis of the criteria of safe shutdown and prevention of significant off-site radiological exposure

  6. Safety culture for engineering companies. Licensing and design bases for Cofrentes NPP

    International Nuclear Information System (INIS)

    Nhorte Gomez, M.D.

    1994-01-01

    Safety culture must be given higher priority by all organisations. It must not be considered a separate concept, attributable to just one particular organisation, or a single responsible party. It is important to apply this criterion throughout the different phases of a nuclear power plant project (design, construction, commissioning and operation) without becoming isolated or dissociated. Nevertheless, it is absolutely essential to apply and consider it during operation, so to ensure highest possible safety standards. Consideration must also be given to the interfaces and interconnections between the different parties involved in the project (Owner of the NPP, Main Engineering Company, Main Supplier, Regulatory Body, etc) to build a SAFETY CULTURE in a collective and effective way. In applying the safety culture, an engineering company emphasises the following concepts: - Personal dedication and sense of responsibility in all those involved in any activity related to the safety of Nuclear Power Plants. - Clearly defined and readily accessible areas of responsibility and channels of communication - Strict adherence to procedures - Internal review of activities (Design review) (Author)

  7. Conceptual Design of a Combined Power Generation Unit at the NPP Seaside

    International Nuclear Information System (INIS)

    Cha, Kyung H.

    2011-01-01

    In order to improve operational performance, an undersea tunnel is being utilized for in-taking and out-taking seawater as coolant in Nuclear Power Plant (NPP). This paper describes a Combined solar-wind-wave Power Generation Unit (CPGU) to be specialized for in-taking and out-taking seawater as coolant in NPP. Accordingly, the purpose of the CPGU is twofold: one is to contain some tunnels to be maintained on the bottom of the CPGU body in order to in-take and out-take coolant water, and the other is to generate a combined power at the NPP seaside. Fig. 1 shows the conceptual CPGU to be configured at the NPP seaside

  8. Relationship towards Engineering, Quality Assurance and 10CFR50.59 in the Design Change Process at the Krsko NPP

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.

    1998-01-01

    The paper discusses the relationship between the Krsko NPP design change engineering practice, quality assurance and the USA Nuclear Regulatory Commission 10 Code Federal Rule 50.59 (10CFR50.59). Together these controls ensure that plant design bases are maintained and yield a safe design. The 10CFR50.59 has been applied in Krsko NPP plant specific procedure entitled ESP-2.303 ''Authorization of Changes, Tests and Experiments'' (Safety Evaluation Screening) since 1994. All proposed changes requiring Safety Evaluations are being submitted to the SNSA (Slovenian Nuclear Safety Administration). If the proposed change is constituting an ''Unreviewed Safety Question'' the formal licensing procedure shall be completed before design change can be implemented otherwise the proposed design change is rejected. The procedure(ESP-2.303) provides the methodology to be followed in determining if a proposed activity involves an unreviewed safety question. An ''Unreviewed Safety Question'' is essentially the same as defined in 10CFR50.59(a)(2): ''A proposed change, test or experiment shall be deemed to involve an unreviewed safety question (1) if the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the (updated) safety analysis report may be increased; or (2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the (updated) safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any technical specification is reduced.'' This paper discusses the Following Krsko NPP Safety Evaluation aspects: 1. Defense in Depth Design Philosophy 2. Methodology 3. Definitions and Applicability of Terms 4. Evaluation Process Guidance and Documentation Process 5. Krsko NPP Lessons Learned. (author)

  9. Main features of buildings and structures important to safety of units V1 and V2 of Bohunice NPP

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

    The program of seismic upgrading of Bohunice NPPs has been started in the year 1989 (after finishing of new seismic input). Since that time the seismic upgrading of Main building of NPP V1 has already been realized, structural as well as technological parts. Beside that the designs of seismic upgrading of other structures of NPP V1 and V2 have been completed. It has been proved that the seismic upgrading of NPPs with reactors WWER 440 is very complicated, but still possible, even in the case with high seismic intensity. It would be not possible to fulfill this complicated task without the help of IAEA Missions. The activities of IAEA experts in the program of Bohunice NPPs upgrading are appreciated very much

  10. Virtual reality in design, planning, operation and training related to the decommissioning of the Chernobyl NPP

    International Nuclear Information System (INIS)

    Mark, N.-K.; Johnsen, T.; Meyer, G.; Owre, F.

    2007-01-01

    Virtual Reality in refueling operation and maintenance training at Leningrad NPP has been recognized by Chernobyl NPP. Institute for Energy Technology's is establishing now the Chernobyl Decommissioning Visualization Centre to be used for planning and training the dismantling procedures in addition to presenting it to the authorities and the public. It will be ready in 2007. The first scenario will be the dismantling of the refueling machines

  11. Seismic hazard, risk, and design for South America

    Science.gov (United States)

    Petersen, Mark D.; Harmsen, Stephen; Jaiswal, Kishor; Rukstales, Kenneth S.; Luco, Nicolas; Haller, Kathleen; Mueller, Charles; Shumway, Allison

    2018-01-01

    We calculate seismic hazard, risk, and design criteria across South America using the latest data, models, and methods to support public officials, scientists, and engineers in earthquake risk mitigation efforts. Updated continental scale seismic hazard models are based on a new seismicity catalog, seismicity rate models, evaluation of earthquake sizes, fault geometry and rate parameters, and ground‐motion models. Resulting probabilistic seismic hazard maps show peak ground acceleration, modified Mercalli intensity, and spectral accelerations at 0.2 and 1 s periods for 2%, 10%, and 50% probabilities of exceedance in 50 yrs. Ground shaking soil amplification at each site is calculated by considering uniform soil that is applied in modern building codes or by applying site‐specific factors based on VS30">VS30 shear‐wave velocities determined through a simple topographic proxy technique. We use these hazard models in conjunction with the Prompt Assessment of Global Earthquakes for Response (PAGER) model to calculate economic and casualty risk. Risk is computed by incorporating the new hazard values amplified by soil, PAGER fragility/vulnerability equations, and LandScan 2012 estimates of population exposure. We also calculate building design values using the guidelines established in the building code provisions. Resulting hazard and associated risk is high along the northern and western coasts of South America, reaching damaging levels of ground shaking in Chile, western Argentina, western Bolivia, Peru, Ecuador, Colombia, Venezuela, and in localized areas distributed across the rest of the continent where historical earthquakes have occurred. Constructing buildings and other structures to account for strong shaking in these regions of high hazard and risk should mitigate losses and reduce casualties from effects of future earthquake strong ground shaking. National models should be developed by scientists and engineers in each country using the best

  12. Seismic design principles for the German fast breeder reactor SNR2

    International Nuclear Information System (INIS)

    Rangette, A.M.; Peters, K.A.

    1988-01-01

    The leading aim of a seismic design is, besides protection against seismic impacts, not to enhance the overall risk in the absence of seismic vibrations and, secondly, to avoid competition between operational needs and a seismic structural design. This approach is supported by avoiding overconservatism in the assumption of seismic loads and in the calculation of the structural response. Accordingly the seismic principles are stated as follows: restriction to German or equivalent low seismicity sites with intensities (SSE) lower VIII at frequency lower than 10 -4 /year; best estimate of seismic input-data without further conservatism; no consideration of OBE. The structural design principles are: 1. The secondary character of the seismic excitation is explicitly accounted for; 2. Energy absorption is allowed for by ductility of materials and construction. Accordingly strain criteria are used for failure predictions instead of stress criteria. (author). 1 fig

  13. Integrated seismic design of structure and control systems

    CERN Document Server

    Castaldo, Paolo

    2014-01-01

    The structural optimization procedure presented in this book makes it possible to achieve seismic protection through integrated structural/control system design. In particular, it is explained how slender structural systems with a high seismic performance can be achieved through inclusion of viscous and viscoelastic dampers as an integral part of the system. Readers are provided with essential introductory information on passive structural control and passive energy dissipation systems. Dynamic analyses of both single and multiple degree of freedom systems are performed in order to verify the achievement of pre-assigned performance targets, and it is explained how the optimal integrated design methodology, also relevant to retrofitting of existing buildings, should be applied. The book illustrates how structural control research is opening up new possibilities in structural forms and configurations without compromising structural performance.

  14. Seismic design practice for Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Chhatre, A.G.; Ingole, S.M.; Bhardwaj, S.A.

    1996-01-01

    Nuclear power plants designed in India in the last twenty years have been designed for earthquake loading using the current licensing practices. Designers and equipment suppliers have therefore been required to consider seismic loading as a major load case. In India, the nuclear power plants have been seismically qualified using state-of-the-art techniques involving both seismic analysis and testing to ensure that the power plant is capable of safely surviving an earthquake that the plant is likely to experience during their operating life. Guidelines and criteria for meeting the qualification requirements are followed as given in various AERB (Indian Atomic Energy Regulatory Board), NRC, IAEA guides, ASME codes and IEEE standards. In this paper various methods available for qualification of structures, systems, mechanical and electrical equipment are explained. The approach and guidelines used within Indian nuclear industry which are evolved from simple analytical requirements to the more elaborate current requirements involving complex analysis and testing on shake table are also summarized

  15. Use of the Human Centered Design concept when designing ergonomic NPP control rooms

    International Nuclear Information System (INIS)

    Skrehot, Petr A.; Houser, Frantisek; Riha, Radek; Tuma, Zdenek

    2015-01-01

    Human-Centered Design is a concept aimed at reconciling human needs on the one hand and limitations posed by the design disposition of the room being designed on the other hand. This paper describes the main aspects of application of the Human-Centered Design concept to the design of nuclear power plant control rooms. (orig.)

  16. Development of an evaluation method for seismic isolation systems of nuclear power facilities. Seismic design analysis methods for crossover piping system

    International Nuclear Information System (INIS)

    Tai, Koichi; Sasajima, Keisuke; Fukushima, Shunsuke; Takamura, Noriyuki; Onishi, Shigenobu

    2014-01-01

    This paper provides seismic design analysis methods suitable for crossover piping system, which connects between seismic isolated building and non-isolated building in the seismic isolated nuclear power plant. Through the numerical study focused on the main steam crossover piping system, seismic response spectrum analysis applying ISM (Independent Support Motion) method with SRSS combination or CCFS (Cross-oscillator, Cross-Floor response Spectrum) method has found to be quite effective for the seismic design of multiply supported crossover piping system. (author)

  17. Seismic design of RC buildings theory and practice

    CERN Document Server

    Manohar, Sharad

    2015-01-01

    This book is intended to serve as a textbook for engineering courses on earthquake resistant design. The book covers important attributes for seismic design such as material properties, damping, ductility, stiffness and strength. The subject coverage commences with simple concepts and proceeds right up to nonlinear analysis and push-over method for checking building adequacy. The book also provides an insight into the design of base isolators highlighting their merits and demerits. Apart from the theoretical approach to design of multi-storey buildings, the book highlights the care required in practical design and construction of various building components. It covers modal analysis in depth including the important missing mass method of analysis and tension shift in shear walls and beams. These have important bearing on reinforcement detailing. Detailed design and construction features are covered for earthquake resistant design of reinforced concrete as well as confined and reinforced masonry structures. Th...

  18. Assessment of seismic hazard for NPP sites in France analysis of several aftershocks of November 8, 1983, Liege earthquake

    International Nuclear Information System (INIS)

    Mohammadioun, B.; Mohammadioun, G.; Bresson, A.

    1984-03-01

    Current French practice for assessing seismic hazard on the sites of nuclear facilities is outlined. The procedure calls for as rich and varied an assortment of actual earthquake recordings as can be procured, including earthquakes in France itself and in nearby countries, recorded by the CEA/IPSN's own staff. Following the November 8, 1983, Liege earthquake, suitably equipped, temporary recording stations were set up in the epicentral area in order to record its aftershocks. Ground motion time histories and response spectra were computed for several of these, and a quality factor Q was derived from these data for the most superficial sedimentary layers of the area. The values obtained show reasonable agreement with ones found for similar materials in other regions

  19. Fast Bayesian Optimal Experimental Design for Seismic Source Inversion

    KAUST Repository

    Long, Quan; Motamed, Mohammad; Tempone, Raul

    2016-01-01

    We develop a fast method for optimally designing experiments [1] in the context of statistical seismic source inversion [2]. In particular, we efficiently compute the optimal number and locations of the receivers or seismographs. The seismic source is modeled by a point moment tensor multiplied by a time-dependent function. The parameters include the source location, moment tensor components, and start time and frequency in the time function. The forward problem is modeled by the elastic wave equations. We show that the Hessian of the cost functional, which is usually defined as the square of the weighted L2 norm of the difference between the experimental data and the simulated data, is proportional to the measurement time and the number of receivers. Consequently, the posterior distribution of the parameters, in a Bayesian setting, concentrates around the true parameters, and we can employ Laplace approximation and speed up the estimation of the expected Kullback-Leibler divergence (expected information gain), the optimality criterion in the experimental design procedure. Since the source parameters span several magnitudes, we use a scaling matrix for efficient control of the condition number of the original Hessian matrix. We use a second-order accurate finite difference method to compute the Hessian matrix and either sparse quadrature or Monte Carlo sampling to carry out numerical integration. We demonstrate the efficiency, accuracy, and applicability of our method on a two-dimensional seismic source inversion problem.

  20. Fast Bayesian optimal experimental design for seismic source inversion

    KAUST Repository

    Long, Quan

    2015-07-01

    We develop a fast method for optimally designing experiments in the context of statistical seismic source inversion. In particular, we efficiently compute the optimal number and locations of the receivers or seismographs. The seismic source is modeled by a point moment tensor multiplied by a time-dependent function. The parameters include the source location, moment tensor components, and start time and frequency in the time function. The forward problem is modeled by elastodynamic wave equations. We show that the Hessian of the cost functional, which is usually defined as the square of the weighted L2 norm of the difference between the experimental data and the simulated data, is proportional to the measurement time and the number of receivers. Consequently, the posterior distribution of the parameters, in a Bayesian setting, concentrates around the "true" parameters, and we can employ Laplace approximation and speed up the estimation of the expected Kullback-Leibler divergence (expected information gain), the optimality criterion in the experimental design procedure. Since the source parameters span several magnitudes, we use a scaling matrix for efficient control of the condition number of the original Hessian matrix. We use a second-order accurate finite difference method to compute the Hessian matrix and either sparse quadrature or Monte Carlo sampling to carry out numerical integration. We demonstrate the efficiency, accuracy, and applicability of our method on a two-dimensional seismic source inversion problem. © 2015 Elsevier B.V.

  1. Fast Bayesian Optimal Experimental Design for Seismic Source Inversion

    KAUST Repository

    Long, Quan

    2016-01-06

    We develop a fast method for optimally designing experiments [1] in the context of statistical seismic source inversion [2]. In particular, we efficiently compute the optimal number and locations of the receivers or seismographs. The seismic source is modeled by a point moment tensor multiplied by a time-dependent function. The parameters include the source location, moment tensor components, and start time and frequency in the time function. The forward problem is modeled by the elastic wave equations. We show that the Hessian of the cost functional, which is usually defined as the square of the weighted L2 norm of the difference between the experimental data and the simulated data, is proportional to the measurement time and the number of receivers. Consequently, the posterior distribution of the parameters, in a Bayesian setting, concentrates around the true parameters, and we can employ Laplace approximation and speed up the estimation of the expected Kullback-Leibler divergence (expected information gain), the optimality criterion in the experimental design procedure. Since the source parameters span several magnitudes, we use a scaling matrix for efficient control of the condition number of the original Hessian matrix. We use a second-order accurate finite difference method to compute the Hessian matrix and either sparse quadrature or Monte Carlo sampling to carry out numerical integration. We demonstrate the efficiency, accuracy, and applicability of our method on a two-dimensional seismic source inversion problem.

  2. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation

  3. A new event detector designed for the Seismic Research Observatories

    Science.gov (United States)

    Murdock, James N.; Hutt, Charles R.

    1983-01-01

    A new short-period event detector has been implemented on the Seismic Research Observatories. For each signal detected, a printed output gives estimates of the time of onset of the signal, direction of the first break, quality of onset, period and maximum amplitude of the signal, and an estimate of the variability of the background noise. On the SRO system, the new algorithm runs ~2.5x faster than the former (power level) detector. This increase in speed is due to the design of the algorithm: all operations can be performed by simple shifts, additions, and comparisons (floating point operations are not required). Even though a narrow-band recursive filter is not used, the algorithm appears to detect events competitively with those algorithms that employ such filters. Tests at Albuquerque Seismological Laboratory on data supplied by Blandford suggest performance commensurate with the on-line detector of the Seismic Data Analysis Center, Alexandria, Virginia.

  4. Preliminary seismic design cost-benefit assessment of the tuff repository waste-handling facilities

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Abrahamson, N.; Hadjian, A.H.

    1989-02-01

    This report presents a preliminary assessment of the costs and benefits associated with changes in the seismic design basis of waste-handling facilities. The objectives of the study are to understand the capability of the current seismic design of the waste-handling facilities to mitigate seismic hazards, evaluate how different design levels and design measures might be used toward mitigating seismic hazards, assess the costs and benefits of alternative seismic design levels, and develop recommendations for possible modifications to the seismic design basis. This preliminary assessment is based primarily on expert judgment solicited in an interdisciplinary workshop environment. The estimated costs for individual attributes and the assumptions underlying these cost estimates (seismic hazard levels, fragilities, radioactive-release scenarios, etc.) are subject to large uncertainties, which are generally identified but not treated explicitly in this preliminary analysis. The major conclusions of the report do not appear to be very sensitive to these uncertainties. 41 refs., 51 figs., 35 tabs

  5. Review of public comments on proposed seismic design criteria

    International Nuclear Information System (INIS)

    Philippacopoulos, A.J.; Shaukat, S.K.; Chokshi, N.C.; Bagchi, G.; Nuclear Regulatory Commission, Washington, DC; Nuclear Regulatory Commission, Washington, DC

    1989-01-01

    During the first quarter of 1988, the Nuclear Regulatory Commission (NRC) prepared a proposed Revision 2 to the NUREG-0800 Standard Review Plan (SRP) Sections 2.5.2 (Vibratory Ground Motion), 3.7.1 (Seismic Design Parameters), 3.7.2 (Seismic Systems Analysis) and 3.7.3 (Seismic Subsystem Analysis). The proposed Revision 2 to the SRP was a result of many years' work carried out by the NRC and the nuclear industry on the Unresolved Safety Issue (USI) A-40: ''Seismic Design Criteria.'' The background material related to NRC's efforts for resolving the A-40 issue is described in NUREG-1233. In June 1988, the proposed Revision 2 of the SRP was issued by NRC for public review and comments. Comments were received from Sargent and Lundy Engineers, Westinghouse Electric Corporation, Stevenson and Associates, Duke Power Company, General Electric Company and Electric Power Research Institute. In September 1988, Brookhaven National Laboratory (BNL) and its consultants (C.J. Costantino, R.P. Kennedy, J. Stevenson, M. Shinozuka and A.S. Veletsos) were requested to carry out a review of the comments received from the above six organizations. The objective of this review was to assist the NRC staff with the evaluation and resolution of the public comments. This review was initiated during October 1988 and it was completed on January 1989. As a result of this review, a set of modifications to the above mentioned sections of the SRP were recommended by BNL and its consultants. This paper summarizes the recommended modifications. 4 refs

  6. Seismic isolation systems designed with distinct multiple frequencies

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1991-01-01

    Two systems for seismic base isolation are presented. The main feature of these system is that, instead of only one isolation frequency as in conventional isolation systems, they are designed to have two distinct isolation frequencies. When the responses during an earthquake exceed the design value(s), the system will automatically and passively shift to the secondly isolation frequency. Responses of these two systems to different ground motions including a harmonic motion with frequency same as the primary isolation frequency, show that no excessive amplification will occur. Adoption of these new systems certainly will greatly enhance the safety and reliability of an isolated superstructure against future strong earthquakes. 3 refs

  7. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    International Nuclear Information System (INIS)

    Shibata, Heki

    1988-01-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  8. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Institute of Industrial Science, University of Tokyo (Japan)

    1988-07-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  9. NPP-Nuclear Island Design. From conceptual design to Project execution

    International Nuclear Information System (INIS)

    Lanchet, Dominique

    2014-01-01

    The second day opened with the lecture of Dominique Lanchet, Design Senior Vice President at AREVA Engineering and Project. Dominique Lanchet gave us an overview of the steps of a Nuclear Island Design creation from the conceptual design to the project execution, giving the examples of the EPR and ATMEA1 TM nuclear reactors

  10. Scope and status of Russian contribution for analysis methods for seismically isolated nuclear structure

    International Nuclear Information System (INIS)

    Beliayev, V.S.; Vinogradov, V.V.; Guskov, V.D.

    1993-01-01

    In the last few years, we can see in Russia the amplification of interest to problems of seismic isolation for potentially dangerous objects as the most effective way to alleviate the possible damage. This material comprises the data which characterize the level of theoretical design and experimental studying of seismic isolation systems of NPP components and structures. (author)

  11. Requirements on PWR reactor design with respect to seismic effects

    International Nuclear Information System (INIS)

    Novak, J.; Pecinka, L.

    1981-01-01

    From the seismic point of view the individual parts of a nuclear power plant must be built such as to allow the shutdown of the reactor up to the safe shutdown earthquake level, the removal of after-heat and the prevention of uncontrolled release of radioactivity into the environment. To the level of operating basic earthquake the plant must be designed such as to allow the operation of the reactor for a period of 100 hours from the seismic event without exceeding the permissible annual dose to personnel and population. The possibility of a loss-of-coolant accident owing to a seismic event is reduced mainly by the integrated performance of the primary circuit, the high-strength structure, the insulation of the main components from the shift of the foundations and the use of floating structures. The pressure vessel of the WWER-1000 reactor is therefore pAaced in a shaft on a support ring and is locked by another support ring. (Z.M.)

  12. 3D seismic isolation for advanced N.P.P application. Hydraulic 3-Dimensional base-isolation system

    International Nuclear Information System (INIS)

    Shimada, Takahiro; Kashiwazaki, Akihiro; Fujiwaka, Tatsuya; Moro, Satoshi

    2003-01-01

    In Japan, a number of three-dimensional base isolation systems have been studied for application to new nuclear plant concepts such as the FBR, but these effects have not so far yielded practically applicable results. The impeding factor has been the difficulty of obtaining an adequate capacity on the vertical isolator for supporting the mass of an actual structure and for suppressing rocking motion. In this paper, we propose a new three-dimensional base isolation system that should solve the foregoing problem. The system is constituted of a set of hydraulic load-carrying cylinders connected to accumulator units containing a compressed gas, another set of rocking-suppression cylinders connected in series, and a laminated rubber bearing laid under each load-carrying cylinder. The present paper covers a basic examination for applying the proposed system to a commercialized FBR now under development in Japan, together with static and dynamic loading tests performed on a scale model to verify expected system performance. Response and analysis reflecting the test results has indicated the proposed system to be well applicable to the envisaged commercialized FBR. The study was undertaken as part of an R and D project sponsored by the government for realizing a three-dimensional seismic isolation system applicable to future FRB's. (author)

  13. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4B. Paks NPP: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on dynamic study of the main building of the Paks NPP; shake table investigation at Paks NPP and the Final report of the Co-ordinated Research Programme

  14. Design on human supervisory control of safety management for advance NPP

    International Nuclear Information System (INIS)

    Nugroho, D.H.; Soentono, S.; Taryo, T.; Wahyon, P.I.

    2006-01-01

    Full text: Full text: Characteristics of an advance NPP related with economic and safety enhancement was represented on capabilities in intelligent control and diagnostic to provide nearly autonomous operation with anticipatory maintenance. An autonomous control system should enable automatic operation while adapting to component faults and system variable upsets. It needs therefore to have many intelligent capabilities, such as modeling, analysis, self-validation, diagnosis and decision. This paper describes a human supervisory control design for nuclear reactor fault management which collaborates between human and autonomous control. The beneficial of collaboration is provided based on belief of information value evaluated from Dempster's rule of evidence. The belief of the collaboration is better compared with single usage. The collaboration was conducted in which agent will autonomously and periodically be conducting surveillance by checking the component abnormalities in the plant if anomalies occur. The anomalies were determined using fault detection module. Thereby the system will be able to conduct preventive maintenance. In the situation of accident happened, hence the system will diagnose to determine the location of component failure autonomously. A human supervisor will then determine the action of decison making based on the prediction result. The decision making will be conducted based on the 4th Sheridan's autonomous level in which the autonomous control will execute the duty autonomously when the plant is in normal condition, or in the predictable accident range. But if the unpredictable accident occurs in the plant, the supervisor will then take over the role to control the plant, and the machine will do what commanded by the supervisor. Sistematically, the system explained before was represented by Traveling Salesman Problem-based surveillance, modified ART-2 artificial neural networks-based fault detection and Bayesian Networks-based fault

  15. Seismic assessment of a site using the time series method

    International Nuclear Information System (INIS)

    Krutzik, N.J.; Rotaru, I.; Bobei, M.; Mingiuc, C.; Serban, V.; Androne, M.

    2001-01-01

    1. To increase the safety of a NPP located on a seismic site, the seismic acceleration level to which the NPP should be qualified must be as representative as possible for that site, with a conservative degree of safety but not too exaggerated. 2. The consideration of the seismic events affecting the site as independent events and the use of statistic methods to define some safety levels with very low annual occurrence probabilities (10 -4 ) may lead to some exaggerations of the seismic safety level. 3. The use of some very high values for the seismic accelerations imposed by the seismic safety levels required by the hazard analysis may lead to very expensive technical solutions that can make the plant operation more difficult and increase the maintenance costs. 4. The consideration of seismic events as a time series with dependence among the events produced may lead to a more representative assessment of a NPP site seismic activity and consequently to a prognosis on the seismic level values to which the NPP would be ensured throughout its life-span. That prognosis should consider the actual seismic activity (including small earthquakes in real time) of the focuses that affect the plant site. The method is useful for two purposes: a) research, i.e. homogenizing the history data basis by the generation of earthquakes during periods lacking information and correlation of the information with the existing information. The aim is to perform the hazard analysis using a homogeneous data set in order to determine the seismic design data for a site; b) operation, i.e. the performance of a prognosis on the seismic activity on a certain site and consideration of preventive measures to minimize the possible effects of an earthquake. 5. The paper proposes the application of Autoregressive Time Series to issue a prognosis on the seismic activity of a focus and presents the analysis on Vrancea focus that affects Cernavoda NPP site by this method. 6. The paper also presents the

  16. 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.

  17. 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.

  18. Researching design solutions for frames of buildings in case of increased seismic intensity in specific zones

    OpenAIRE

    Panasyuk Leonid; Kravchenko Galina; Trufanova Elena

    2017-01-01

    Currently, there is a trend to increase the estimated seismic hazard for construction sites. With this, the buildings erected under the previously valid norms have the lesser hazard resistance. The present article inquiries into an issue of how the design solutions affect the safety of the building change under the increased seismic intensity. This article represents the calculation of a building without regard to seismic intensity and the same was made for a rate-7 seismic intensity district...

  19. Seismic design criteria of fire protection systems for DOE facilities

    International Nuclear Information System (INIS)

    Hardy, G.; Cushing, R.; Driesen, G.

    1991-01-01

    Fire protection systems are critical to the safety of personnel and to the protection of inventory during any kind of emergency situation that involves a fire. The importance of these fire protection systems is hightened for DOE facilities which often house nuclear, chemical or scientific processes. Current research into the topic of open-quotes fires following earthquakesclose quotes has demonstrated that the risks of a fire starting as a result of a major earthquake can be significant. Thus, fire protection systems need to be designed to withstand the anticipated seismic event for the site in question

  20. The design preparation for radiation monitoring system in the frame of completion NPP Mochovce Units 3 and 4

    International Nuclear Information System (INIS)

    Sevecka, S.; Slavik, O.; Kapisovsky, V.

    2009-01-01

    In 1985 a Basic Design of Radiation Monitoring System (RMS) has been elaborated for Mochovce NPP unit 3 and 4 construction. Due to construction interruption in the following years this design solution became obsolete. A new solution of RMS have been developed with conception following that of original Basic Design accommodating also safety measures implemented in RMS of NPP EMO units 1 and 2, and based on modem instrumentation and computer technique. Following the updating of Basic Design documentation the preparation of elaboration of RMS detailed design was carried on. In the frame of this preparation a review of possible suppliers of instrumentation satisfying the conception of radiation monitoring system and the extension of required deliveries has been made. Also criteria on RMS suppliers selection have been determined. The types of monitoring systems and equipment, as well as their quantities, have been specified based on updated Basic Design requirements and production profiles and possibilities of potential suppliers. The required parameters of measurements (including measurement geometry) have been evaluated, as well as requirements of legislation and requirements of proposed RMS architecture. (authors)

  1. Exploratory Shaft Seismic Design Basis Working Group report

    International Nuclear Information System (INIS)

    Subramanian, C.V.; King, J.L.; Perkins, D.M.; Mudd, R.W.; Richardson, A.M.; Calovini, J.C.; Van Eeckhout, E.; Emerson, D.O.

    1990-08-01

    This report was prepared for the Yucca Mountain Project (YMP), which is managed by the US Department of Energy. The participants in the YMP are investigating the suitability of a site at Yucca Mountain, Nevada, for construction of a repository for high-level radioactive waste. An exploratory shaft facility (ESF) will be constructed to permit site characterization. The major components of the ESF are two shafts that will be used to provide access to the underground test areas for men, utilities, and ventilation. If a repository is constructed at the site, the exploratory shafts will be converted for use as intake ventilation shafts. In the context of both underground nuclear explosions (conducted at the nearby Nevada Test Site) and earthquakes, the report contains discussions of faulting potential at the site, control motions at depth, material properties of the different rock layers relevant to seismic design, the strain tensor for each of the waveforms along the shaft liners, and the method for combining the different strain components along the shaft liners. The report also describes analytic methods, assumptions used to ensure conservatism, and uncertainties in the data. The analyses show that none of the shafts' structures, systems, or components are important to public radiological safety; therefore, the shafts need only be designed to ensure worker safety, and the report recommends seismic design parameters appropriate for this purpose. 31 refs., 5 figs., 6 tabs

  2. Innovative design of viscoelastic dampers for seismic mitigation

    International Nuclear Information System (INIS)

    Tsai, C.S.

    1993-01-01

    In this paper, an advanced and more reliable design of viscoelastic dampers for seismic mitigation of high-rise buildings is presented. The innovative design of energy-absorbing devices has some advantages, compared to the classical design, as follows: One, the device is directly subjected to shear strains and forces due to story drifts; two, the device can support its own weight during normal operations, and maintain stable for large deformations during earthquakes; three, the device can reduce the responses of a structure to horizontal as well as vertical seismic loadings; and four, the device can also decrease the responses of the floor system of a building. In this study, a ten-story building is given as an example to express the merits obtained from the new system. Comparisons of the building equipped with classical and proposed devices of viscoelastic dampers are carefully studied. Numerical results show that the energy-absorbing capacity of the new device is superior to the classical one, especially for vertical vibrations. (orig.)

  3. Final report of the cooperative study on seismic isolation design. The second stage

    Energy Technology Data Exchange (ETDEWEB)

    Uryu, Mitsuru; Terada, Syuji; Shioya, Tsutomu (and others)

    1999-05-01

    The applicability of the seismic isolation design onto the nuclear fuel facilities, which must clear severe criteria of integrity, has been examined. Following the first stage of the cooperative study, conducted from 1988 to 1991, the second stage included critical vibration testing, seismic observation of seismic isolation building and founded buildings of non-isolation, with the objectives of clarifying the policies on critical design of seismic isolation building. Integrity of the seismic isolation piping system was tested by means of static deformation test, with variable inner water pressure and relative deformation. (Yamamoto, A.)

  4. Seismic design criteria and their application to major hazard plant within the United Kingdom

    International Nuclear Information System (INIS)

    Alderson, M.A.H.G.

    1982-12-01

    The nature of seismic motions and the implications are briefly described and the development of seismic design criteria for nuclear power plants in various countries is described including possible future developments. The seismicity of the United Kingdom is briefly reviewed leading to the present position on seismic design criteria for nuclear power plants within the United Kingdom. Damage from past destructive earthquakes is reviewed and the existing codes of practice and standards are described. Finally the effect of earthquakes on major hazard plant is discussed in general terms including the seismic analysis of a typical plant item. (author)

  5. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4D. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs; structural analysis and site inspection for site requalification; structural response of Paks NPP reactor building; analysis and testing of model worm type tanks on shaking table; vibration test of a worm tank model; evaluation of potential hazard for operating WWER control rods under seismic excitation

  6. Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

    2012-03-15

    JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

  7. Seismic resistant design of a nuclear category I earth dam

    International Nuclear Information System (INIS)

    Vaidya, N.R.; Ries, E.R.; Kissenpfennig, J.F.

    1975-01-01

    An integral part of many nuclear power plants is the ultimate heat sink (UHS); the purpose of which is to retain and deliver a supply of service water to the plant when water from the primary circulating water system is not available. The earth dam described herein is designed to retain the reservoir for the UHS of a nuclear power plant in Southern Europe. The usual pseudo-static analysis is only as good as the estimate for the seismic coefficient used to compute an equivalent horizontal static force on a potential sliding mass. In view of the earth dam considered herein, a more accurate computation of the seismic coefficients is to be made. A two-dimensional dynamic finite element analysis is made to predict the response of the earth dam to a Safe Shutdown Earthquake excitation which is in the form of a time history of accelerations appropriately deconvoluted from the surficial time history and applied at the base of the model. The material properties such as shear modulus and damping are adjusted to be compatible with the level of strain obtained. Thus, non-linear behavior of soil is considered in the analysis and a more realistic response is predicted. Acceleration and stress are determined throughout the dam and are used to compute a seismic coefficient for a pseudo-static stability analysis and the dynamic strength to stress ratios at several points in the body of the dam. The need to design the dam to resist a progressive erosion accident resulting from postulated concentrated leaks is discussed. This may be accomplished by providing a wide, well graded core protected by wide transition cores also heavily compacted

  8. Sitting Safety Aspects of Second Romanian NPP

    International Nuclear Information System (INIS)

    Mauna, T.

    2010-01-01

    The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units like the Wolsong applied design project for nuclear island. For the BOP parts the ASALDO-GE project was applied with the careful about the interface connection NSP requirements. The new NPP sitting studies began from 1982 in a serious manner as first part on Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. For develop the all package of the studies in concordance with the first IAEA Safety Standards recommendations. Till the 1982 the first mission of design and research multi-branch of specialists team was to adapt the NPP Cernavoda project having a open water cooling circuit to the new parameters of close water cooling circuit. But the team was looking at the other type of NPP for sitting. Also in the same time was studied the possibility of NSP foundation on hard less or soft soil foundation strata in connection with safety aspects. The close circuit of cooling water means others parameters of systems and need very large cooling towers. Also must be reconsidering the safety systems design and performance as new solution. In the south of Transylvania historical region in Romania the Olt River run from west to east having medium multi annual flow around 70 m3/s. The Olt River has a chain of small hydropower in operation and other planned. From geological and geophysical points of view two main faults, along the Olt river valley, one of this having seismically small activities was detected. Site region geotechnical studies show small quantity underground natural gas, salt and peat. The initial nuclear program has imposed 4 NPP units site near Olt River. Taking into account the orogenesis, water cooling needs and other local feature can't be built more than two NPP units on a site. This paper tries to reconsider the old analysis from the last IAEA Safety Standards point of view taking into account the new

  9. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  10. Seismic design of a uranium conversion plant building

    International Nuclear Information System (INIS)

    Peixoto, O.J.M.; Botelho, C.L.A.; Braganca, A. Jr.; C. Santos, S.H. de.

    1992-01-01

    The design of facilities with small radioactive inventory has been traditionally performed following the usual criteria for industrial buildings. In the last few years, more stringent criteria have been adopted in new nuclear facilities in order to achieve higher standards for environmental protection. In uranium conversion plants, the UF 6 (uranium hexafluoride) production step is the part of the process with the highest potential for radioactivity release to the environment because of the operations performed in the UF 6 desublimers and cylinder filling areas as well as UF 6 distillation facilities, when they are also required in the process. This paper presents the design guidelines and some details of the seismic resistance design of a UF 6 production building to be constructed in Brazil

  11. Study of seismic design bases and site conditions for nuclear power plants

    International Nuclear Information System (INIS)

    1980-04-01

    This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches

  12. Study of seismic design bases and site conditions for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    This report presents the results of an investigation of four topics pertinent to the seismic design of nuclear power plants: Design accelerations by regions of the continental United States; review and compilation of design-basis seismic levels and soil conditions for existing nuclear power plants; regional distribution of shear wave velocity of foundation materials at nuclear power plant sites; and technical review of surface-founded seismic analysis versus embedded approaches.

  13. The challenge of the global management of plant design modifications. example of the new EJ system at Vandellos NPP

    International Nuclear Information System (INIS)

    Ortega, Fernando; Valdivia, Carlos; Fernandez Illobre, Luis; Trueba, Pedro

    2010-01-01

    One of the most challenging areas in the operation of nuclear power plants (NPP) is related to the management of plant design modifications. Plant modifications can be made to improve reliability, facilitate operation, improve safety or get better results. In any of these situations, plant modifications imply many different activities that have to be done in a coordinated manner. NUREG-0711 (Human Factors Engineering Program Review Model) shows a global approach to manage most of these activities. Although this approach is mainly focused on the design and construction of new plants, it can also be applied to plant modification management. Successful global management will require performing every activity in a specific order, taking advantage of the output coming from some tasks as input for others and finalizing every task when necessary. This will provide the best results in terms of quality, time required for implementation, safe and reliable operation and maintenance, and cost. Tecnatom is involved in most of the activities related to the operational areas and has applied a global approach to get advantages in terms of quality and cost, which is outlined in this paper. As an example of this approach, the Vandellos NPP experience is shown in this presentation. Vandellos NPP carried out an important design modification that consists of replacing an old essential service water system with a new one. This was a three-year project that implied the construction of new reservoirs, new buildings, the implementation of new equipment, and new panels in the main control room. This paper shows the way in which all of these activities were performed. (authors)

  14. A Survey study on design procedure of Seismic Base Isolation ...

    African Journals Online (AJOL)

    Michael Horsfall

    Base Isolation Systems that is flexible approach to decrease the potential damage. In this ... In addition, we analyze the seismic responses of isolated structures. The seismic ..... Equation 3.7, is examined; it is realized that the inequality ...

  15. Studies on the seismic buckling design guideline of FBR main vessels. 9. Buckling evaluation under elastic-plastic seismic response

    International Nuclear Information System (INIS)

    Hagiwara, Yutaka; Yamamoto, Kohsuke; Kawamoto, Yoji; Nakagawa, Masaki; Akiyama, Hiroshi

    1998-01-01

    Plastic shear-bending buckling under seismic loadings is one of the major problems in the structural design of FBR main vessels. Pseudo-dynamic and dynamic buckling tests of cylinders were performed in order to study the effects of nonlinear seismic response on buckling strength, ductility, and plastic response reduction. The buckling strength formulae and the rule for ductility factors both derived from static tests were confirmed to be valid for the tests under dynamic loads. The displacement-constant rule for response reduction effect was modified by acceleration amplification factor in order to maintain applicability for various spectral profiles of seismic excitations. The response reduction estimated by the proposed rule was reasonably conservative for all cases of the pseudo-dynamic and the dynamic tests. Finally, a seismic safety assessment rule was proposed for plastic shear-bending buckling of cylinders, which include the proposed response reduction rule. (author)

  16. Endurance time method for Seismic analysis and design of structures

    International Nuclear Information System (INIS)

    Estekanchi, H.E.; Vafai, A.; Sadeghazar, M.

    2004-01-01

    In this paper, a new method for performance based earthquake analysis and design has been introduced. In this method, the structure is subjected to accelerograms that impose increasing dynamic demand on the structure with time. Specified damage indexes are monitored up to the collapse level or other performance limit that defines the endurance limit point for the structure. Also, a method for generating standard intensifying accelerograms has been described. Three accelerograms have been generated using this method. Furthermore, the concept of Endurance Time has been described by applying these accelerograms to single and multi degree of freedom linear systems. The application of this method for analysis of complex nonlinear systems has been explained. Endurance Time method provides a uniform approach to seismic analysis and design of complex structures that can be applied in numerical and experimental investigations

  17. Development of rational design technique for frame steel structure combining seismic resistance and economic performance

    International Nuclear Information System (INIS)

    Kato, Motoki; Morishita, Kunihiro; Shimono, Masaki; Chuman, Yasuharu; Okafuji, Takashi; Monaka, Toshiaki

    2015-01-01

    Anti-seismic designs have been applied to plant support steel frames for years. Today, a rational structure that further improves seismic resistance and ensures economic performance is required in response to an increase of seismic load on the assumption of predicted future massive earthquakes. For satisfying this requirement, a steel frame design method that combines a steel frame weight minimizing method, which enables economic design through simultaneous minimization of multiple steel frame materials, and a seismic response control design technology that improves seismic resistance has been established. Its application in the design of real structures has been promoted. This paper gives an overview of this design technology and presents design examples to which this design technology is applied. (author)

  18. Determination of Design Basis Earthquake ground motion

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1997-01-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  19. Determination of Design Basis Earthquake ground motion

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Muneaki [Japan Atomic Power Co., Tokyo (Japan)

    1997-03-01

    This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

  20. 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.)

  1. 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.)

  2. Japan Catastrophic Earthquake and Tsunami in Fukushima Daiichi NPP; Is it Beyond Design Basis Accident or a Design Deficiency and Operator Unawareness?

    International Nuclear Information System (INIS)

    Gaafar, M.A.; Refeat, R.M.; EL-Kady, A.A.

    2012-01-01

    On March 11, 2011 a catastrophic earthquake and tsunami struck the north east coast of Japan. This catastrophe damaged fully or partially the six units of the Fukushima Daiichi Nuclear Power Plant.Questions were raised following the aftermath, whether it is beyond design basis accident caused by severe natural event or a failure by the Japanese authorities to plan to deal with such accident. There are many indications that the Utility of Fukushima Daiichi NPP, Tokyo Electric Power Company (TEPCO), did not pay enough attention to numerous facts about the incompatibility of the site and several design defects in the plant units. In fact there are three other NPP sites nearby Fukushima Daiichi Plant (about 30 to 60 Km far from Fukushima Daiichi NPP), with different site characteristics, which survived the same catastrophic earthquake and tsunami, but they were automatically turned into a safe shutdown state. These plants sites are Fukushima Daini Plant (4 units), Onagawa Plant (3 units) and Tokai Daini (II) Plant (one unit). In this paper, the aftermath Fukushima Daiichi plant integrity is pointed out. Some facts about the site and design concerns which could have implications on the accident are discussed. The response of Japan Authority is outlined and some remarks about their actions are underlined. The impacts of this disaster on the Nuclear Power Program worldwide are also discussed.

  3. Seismic behavior and design of wall-EDD-frame systems

    Directory of Open Access Journals (Sweden)

    Oren eLavan

    2015-06-01

    Full Text Available Walls and frames have different deflection lines and, depending on the seismic mass they support, may often poses different natural periods. In many cases, wall-frame structures present an advantageous behavior. In these structures the walls and the frames are rigidly connected. Nevertheless, if the walls and the frames were not rigidly connected, an opportunity for an efficient passive control strategy would arise: Connecting the two systems by energy dissipation devices (EDDs to result in wall-EDD-frame systems. This, depending on the parameters of the system, is expected to lead to an efficient energy dissipation mechanism.This paper studies the seismic behavior of wall-EDD-frame systems in the context of retrofitting existing frame structures. The controlling non-dimensional parameters of such systems are first identified. This is followed by a rigorous and extensive parametric study that reveals the pros and cons of the new system versus wall-frame systems. The effect of the controlling parameters on the behavior of the new system are analyzed and discussed. Finally, tools are given for initial design of such retrofitting schemes. These enable both choosing the most appropriate retrofitting alternative and selecting initial values for its parameters.

  4. Design of the Caltrans Seismic Response Modification Device (SRMD) test facility

    International Nuclear Information System (INIS)

    Benzoni, G.; Seible, F.

    1998-01-01

    In the Seismic retrofit design of California's Toll Bridges, seismic isolation is used in several bridges to limit the seismic force input into the superstructure and to avoid costly superstructure retrofit measures which would require partial lane closures and traffic interruptions. Isolation bearings and dampers of the size required for these large span bridges have not been built or tested to date. This paper describes the design and construction of a full scale testing facility which will allow the real-time 6-DOF dynamic characterization of the seismic response modification devices designed for California's Toll Bridges. (author)

  5. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  6. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  7. Behaviour of NPP Kozloduy (Bulgaria) during the three strong Vranchea earthquakes

    International Nuclear Information System (INIS)

    Sachanski, S.; Krutzik, N.; Sachanski, G.

    1991-01-01

    Following the new tendency a coordinated program with IAEA-Vienna for increasing of NPP Kozloduy seismic safety is in realisation. To check the real characteristics of 440 MW blocks, to establish more realistic mathematical model and response of structure is purpose of this study. The revaluated and increased the design peak ground acceleration with 40 % needs additional analysis of the structures and equipment. Space mathematical model and three component seismic excitation, represented with power spectra density function, are under consideration for revaluation of seismic safety. (author)

  8. Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

    International Nuclear Information System (INIS)

    Hinovski, M.P.; Sabinov, S.

    2001-01-01

    Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

  9. Seismic design criteria for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Morrone, A.; Bitner, J.L.; Sigal, G.B.

    1975-01-01

    The general criteria for seismic resistant design for structures, systems and components of the Clinch River Breeder Reactor Plant (CRBRP) are presented and discussed. Site dependency of the maximum ground accelerations for the Operating Basis Earthquake and the Safe Shutdown Earthquake is described from the viewpoint of historical records and geological and seismological studies for the CRBRP site. The respective ground response spectra are derived by normalization of the latest AEC Regulatory standard shapes to these maximum ground accelerations. Modeling and analytical techniques and requirements are given. In addition, loading conditions and categories, loading combinations, earthquake direction effects and allowable damping values are defined. A discussion of the testing criteria which considers both single and multiple frequency test motions, and basic test procedures for single frequency sine beat testing is presented. (U.S.)

  10. Calculation of anti-seismic design for Xi'an pulsed reactor

    International Nuclear Information System (INIS)

    Li Shuian

    2002-01-01

    The author describes the reactor safety rule, safety regulation and design code that must be observed to anti-seismic design in Xi'an pulsed reactor. It includes the classification of reactor installation, determination of seismic loads, calculate contents, program, method, results and synthetically evaluation. According to the different anti-seismic structure character of reactor installation, an appropriate method was selected to calculate the seismic response. The results were evaluated synthetically using the design code and design requirement. The evaluate results showed that the anti-seismic design function of reactor installation of Xi'an pules reactor is well, and the structure integrality and normal property of reactor installation can be protect under the designed classification of the earthquake

  11. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

    Energy Technology Data Exchange (ETDEWEB)

    1985-04-01

    This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.

  12. Seismic design of nuclear power plants - where are we now?

    International Nuclear Information System (INIS)

    Roesset, J.M.

    1998-01-01

    The lack of any significant activity in the design and construction of new nuclear power plants over the last 10 years has resulted in a corresponding lull in the basic academic research carried out in this field. Whilst some work is still going on related to the evaluation of existing plants or to litigation over some of them (including some that never became operational) most of it is of a very applied nature and little basic research is being conducted at present. However, research on earthquake engineering in general, as applied to buildings, bridges, lifelines, dams and other constructed facilities has continued. This paper attempts to look at some of the areas where there were major uncertainties in the seismic design of nuclear power plants (selection of the design earthquake and its characteristics, evaluation of soil effects and soil structure interactions, dynamic analysis and design of the structures), the progress that has been made in these areas, and the remaining issues in need of further research. (orig.)

  13. Seismic design of nuclear power plants. Where are we now?

    International Nuclear Information System (INIS)

    Roesset, J.M.

    1995-01-01

    The lack of any significant activity in the design and construction of new nuclear power plants over the last ten years has resulted in a corresponding lull in the basic academic research carried out in this field. While some work is still going on related to the evaluation of existing plants or to litigation over some of them (including some that never became operational) most of it is of a very applied nature and little basic research is being conducted at present. Yet research on earthquake engineering in general, as applied to buildings, bridges, lifelines, dams and other constructed facilities has continued. This paper attempts to look at some of the areas where there were major uncertainties in the seismic design of nuclear power plants (selection of the design earthquake and its characteristics, evaluation of soil effects and soil structure interactions, dynamic analysis and design of the structures), the progress that has been made in these areas, and the remaining issues in need of further research. (author)

  14. Latest standards on seismic resistance related to research activities

    International Nuclear Information System (INIS)

    Juhasova, Emilia

    2002-01-01

    The paper discus few basic approaches applied in final drafts of prEN 1990 and prEN 1998-1. It is pointed out on design working life, loads combinations and the range of behaviour factors for concrete, steel and masonry buildings. The procedure and main results of large masonry model seismic tests are presented. As far as the masonry walls create the part of many NPP structures obtained results could be utilised also for the increase of their seismic resistance

  15. Upgrading accuracy of designed seismic vibration on concept of the land conditions

    International Nuclear Information System (INIS)

    Tamura, Keichi; Kaneko, Masahiro; Honda, Toshiki; Chiba, Hikaru

    1998-01-01

    In this study, some investigations on design procedure of designed seismic vibration were conducted on concept of amplification of the seismic vibration and nonlinearity of the system at the place largely changing topographic and land conditions. In this fiscal year, after collecting and arranging the topographic and land conditions at settling place of the nuclear facilities and their circumferences, some investigations on effect of the seismic vibration amplified at surface layer of grounds on behavior of nonlinear system as well as arrangement of relationship between the topographic and land conditions and seismic vibration amplifying properties at the surface layer of grounds were conducted. (G.K.)

  16. Assessment of the impact of degraded shear wall stiffnesses on seismic plant risk and seismic design loads

    International Nuclear Information System (INIS)

    Klamerus, E.W.; Bohn, M.P.; Johnson, J.J.; Asfura, A.P.; Doyle, D.J.

    1994-02-01

    Test results sponsored by the USNRC have shown that reinforced shear wall (Seismic Category I) structures exhibit stiffnesses and natural frequencies which are smaller than those calculated in the design process. The USNRC has sponsored Sandia National Labs to perform an evaluation of the effects of the reduced frequencies on several existing seismic PRAs in order to determine the seismic risk implications inherent in these test results. This report presents the results for the re-evaluation of the seismic risk for three nuclear power plants: the Peach Bottom Atomic Power Station, the Zion Nuclear Power Plant, and Arkansas Nuclear One -- Unit 1 (ANO-1). Increases in core damage frequencies for seismic initiated events at Peach Bottom were 25 to 30 percent (depending on whether LLNL or EPRI hazard curves were used). At the ANO-1 site, the corresponding increases in plant risk were 10 percent (for each set of hazard curves). Finally, at Zion, there was essentially no change in the computed core damage frequency when the reduction in shear wall stiffness was included. In addition, an evaluation of deterministic ''design-like'' structural dynamic calculations with and without the shear stiffness reductions was made. Deterministic loads calculated for these two cases typically increased on the order of 10 to 20 percent for the affected structures

  17. Views on seismic design standardization of structures, systems and components of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.

    2011-01-01

    Structures, Systems and Components (SSCs) of nuclear facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Manmade accidents such as aircraft impact, explosions etc., sometimes may be considered as design basis event and sometimes taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event which has certain annual frequency specified in design codes. For example nuclear power plants are designed for a seismic event has 10000 year return period. It is generally felt that design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to

  18. Refuelling design and core calculations at NPP Paks: codes and methods

    International Nuclear Information System (INIS)

    Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

    2001-01-01

    This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

  19. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of); Choi, Young Jae; Kim, Yong Jin [KAIST, Daejeon (Korea, Republic of); Park, Hyo Chan; Park, Youn Won [BEES, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now.

  20. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    International Nuclear Information System (INIS)

    Song, Sub Lee; Choi, Young Jae; Kim, Yong Jin; Park, Hyo Chan; Park, Youn Won

    2016-01-01

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now

  1. TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

    International Nuclear Information System (INIS)

    Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

    2010-01-01

    TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

  2. Design and analysis of fractional order seismic transducer for displacement and acceleration measurements

    Science.gov (United States)

    Veeraian, Parthasarathi; Gandhi, Uma; Mangalanathan, Umapathy

    2018-04-01

    Seismic transducers are widely used for measurement of displacement, velocity, and acceleration. This paper presents the design of seismic transducer in the fractional domain for the measurement of displacement and acceleration. The fractional order transfer function for seismic displacement and acceleration transducer are derived using Grünwald-Letnikov derivative. Frequency response analysis of fractional order seismic displacement transducer (FOSDT) and fractional order seismic acceleration transducer (FOSAT) are carried out for different damping ratio with the different fractional order, and the maximum dynamic measurement range is identified. The results demonstrate that fractional order seismic transducer has increased dynamic measurement range and less phase distortion as compared to the conventional seismic transducer even with a lower damping ratio. Time response of FOSDT and FOSAT are derived analytically in terms of Mittag-Leffler function, the effect of fractional behavior in the time domain is evaluated from the impulse and step response. The fractional order system is found to have significantly reduced overshoot as compared to the conventional transducer. The fractional order seismic transducer design proposed in this paper is illustrated with a design example for FOSDT and FOSAT. Finally, an electrical equivalent of FOSDT and FOSAT is considered, and its frequency response is found to be in close agreement with the proposed fractional order seismic transducer.

  3. Engineering Seismic Base Layer for Defining Design Earthquake Motion

    International Nuclear Information System (INIS)

    Yoshida, Nozomu

    2008-01-01

    Engineer's common sense that incident wave is common in a widespread area at the engineering seismic base layer is shown not to be correct. An exhibiting example is first shown, which indicates that earthquake motion at the ground surface evaluated by the analysis considering the ground from a seismic bedrock to a ground surface simultaneously (continuous analysis) is different from the one by the analysis in which the ground is separated at the engineering seismic base layer and analyzed separately (separate analysis). The reason is investigated by several approaches. Investigation based on eigen value problem indicates that the first predominant period in the continuous analysis cannot be found in the separate analysis, and predominant period at higher order does not match in the upper and lower ground in the separate analysis. The earthquake response analysis indicates that reflected wave at the engineering seismic base layer is not zero, which indicates that conventional engineering seismic base layer does not work as expected by the term ''base''. All these results indicate that wave that goes down to the deep depths after reflecting in the surface layer and again reflects at the seismic bedrock cannot be neglected in evaluating the response at the ground surface. In other words, interaction between the surface layer and/or layers between seismic bedrock and engineering seismic base layer cannot be neglected in evaluating the earthquake motion at the ground surface

  4. On seismic design of cable trays and their supports

    International Nuclear Information System (INIS)

    Hartmann, B.

    1978-01-01

    Codes presently in force for design of nuclear power plants require seismic qualification for all electric equipment. In the case of cable trays and their supports one usually attempts to meet the requirements of the code by stiffening a standardized design. This procedure leads to impracticall,imensions for the mountings and, above all, to the loss of the modular character. With strong earthquakes however, it may become irrational at all. This paper suggests an alternate strategy. It starts with a standardized system again, adding some units. These are on the one hand diagonal bracing elements, arbitrarily to arrange, thus gaining a more or less rigid supporting framework. And on the other hand as an essential modification, elastomer rubber pads are inserted as spring bearings. With these pads between the supporting and the adjoining structure, the assembly becomes tractable with respect to earthquake qualification. The question of material properties is also addressed. The elastomer pads have to be chosen so as to fulfil all expected functions under usual as well as extreme environmental conditions. (Author)

  5. Development of a HFE program for an operating NPP: Balancing between existing design practices and human factors standards

    International Nuclear Information System (INIS)

    Salo, Leena; Savioja, Paula

    2014-01-01

    This paper describes HFE program development project conducted at a Finnish power company Fortum. The aim of developing a formal HFE program was to improve integration of human factors issues in design of technical systems and to systematically document the HFE process of the company. As Fortum has a long tradition of designing control room solutions, the starting point of the HFE program development was the existing own design practices. On the other hand, the aim was to create a program which would comply with international guidelines such as NUREG-0711. The program development was conducted by tracing the HFE design practices in an on-going I and C modernization project. This empirical work was carried out by interviews of designers and other HFE key stake holders. After the explication of the current practices, the gaps, overlaps and differences in relation to the international standards and guidelines were identified. Based on an analysis of current practices and guidelines and standards a new HFE process model was created. The design process model can be followed in modifications which concern systems with human user interfaces of any kind. The model consists of five separate phases which comply with the general engineering design process model utilized at the company. The HFE program is intended to be both a practical guide on how to take human factors issues into consideration in the design of NPP systems and also a tool for the management of HFE activities

  6. Structure design of human factor data management system for Daya Bay NPP

    International Nuclear Information System (INIS)

    Zhang Li; Zhang Ning; Guo Jianbing; Huang Weigang; Zhu Minhong; Wang Jin

    2000-01-01

    Collection, analysis and quantification of human factor data are important compositions of human reliability analysis (HRA) and probabilistic risk assessment (PRA). Various human factor databases have been set up, but there are comparatively little human factor data management systems which can be uses for collection, classification, analysis, calculation and predication of the human factor data. Therefore, the human factor data management system for Daya Bay NPP is developed, with the following three modules and four databases: original data module, computing module, introduced data module, and basic database, other data source of the plant, external database and introduced database. The foundational problems about human factor data and the systemic structure and function are described. The data structure in the database is also discussed, because it is of the most importance in the system

  7. Towards Improved Considerations of Risk in Seismic Design (Plinius Medal Lecture)

    Science.gov (United States)

    Sullivan, T. J.

    2012-04-01

    The aftermath of recent earthquakes is a reminder that seismic risk is a very relevant issue for our communities. Implicit within the seismic design standards currently in place around the world is that minimum acceptable levels of seismic risk will be ensured through design in accordance with the codes. All the same, none of the design standards specify what the minimum acceptable level of seismic risk actually is. Instead, a series of deterministic limit states are set which engineers then demonstrate are satisfied for their structure, typically through the use of elastic dynamic analyses adjusted to account for non-linear response using a set of empirical correction factors. From the early nineties the seismic engineering community has begun to recognise numerous fundamental shortcomings with such seismic design procedures in modern codes. Deficiencies include the use of elastic dynamic analysis for the prediction of inelastic force distributions, the assignment of uniform behaviour factors for structural typologies irrespective of the structural proportions and expected deformation demands, and the assumption that hysteretic properties of a structure do not affect the seismic displacement demands, amongst other things. In light of this a number of possibilities have emerged for improved control of risk through seismic design, with several innovative displacement-based seismic design methods now well developed. For a specific seismic design intensity, such methods provide a more rational means of controlling the response of a structure to satisfy performance limit states. While the development of such methodologies does mark a significant step forward for the control of seismic risk, they do not, on their own, identify the seismic risk of a newly designed structure. In the U.S. a rather elaborate performance-based earthquake engineering (PBEE) framework is under development, with the aim of providing seismic loss estimates for new buildings. The PBEE framework

  8. 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

  9. Differences in safety margins between nuclear and conventional design standards with regards to seismic hazard definition and design criteria

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Orbovic, N.; Dejan, D.

    2006-01-01

    With the surging interest in new build nuclear all over the world and a permanent interest in earthquake resistance of nuclear plants, there is a need to quantify the safety margins in nuclear buildings design in comparison to conventional buildings in order to increase the public confidence in the safety of nuclear power plants. Nuclear (CAN3-N289 series) and conventional (NBCC 2005) seismic standards have different approaches regarding the design of civil structures. The origin of the differences lays in the safety philosophy behind the seismic nuclear and conventional standards. Conventional seismic codes contain the minimal requirement destined primarily to safeguard against major structural failure and loss of life. It doesn't limit damage to a certain acceptable degree or maintain function. Nuclear seismic code requires that structures, systems and components important to safety, withstand the effects of earthquakes. The requirement states that for equipment important to safety, both integrity and functionality should be ascertained. The seismic hazard is generally defined on the basis of the annual probability of exceedence (return period). There is a major difference on the return period and the confidence level for design earthquakes between the conventional and the nuclear seismic standards. The seismic design criteria of conventional structures are based on the use of Force Modification Factors to take into account the energy dissipation by incursion in non-elastic domain and the reserve of strength. The use of such factors to lower intentionally the seismic input is consistent with the safety philosophy of the conventional seismic standard which is the 'non collapse' rather than the integrity and/or the operability of the structures or components. Nuclear seismic standard requires that the structure remain in the elastic domain; energy dissipation by incursion in non-elastic domain is not allowed for design basis earthquake conditions. This is

  10. Seismic analysis, evaluation and upgrade design for a DOE exhaust stack building

    International Nuclear Information System (INIS)

    Malik, L.E.; Maryak, M.E.

    1991-01-01

    An exhaust stack building of a nuclear reactor facility with complex structural configuration has been analyzed and evaluated and retrofitted for seismic forces. The building was built in the 1950's and had not been designed to resist seismic forces. A rigorous analysis and evaluation program was implemented to minimize costly retrofits required to upgrade the building to resist high seismic forces. Seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, basemat flexibility and the influence of the nearby reactor building were considered in rigorous seismic analyses. These analyses and evaluations enabled limited upgrades to qualify the stack building for the seismic forces. Some of the major conclusions of this study are: (1) a phased approach of seismic analyses, utilizing simplified models to evaluate practicable upgrade schemes, and, then incorporating the most suitable scheme in a rigorous model to obtain design forces for upgrades, is an efficient and cost-effective approach for seismic qualification of nuclear facilities to higher seismic criteria; and, (2) finalizing the upgrade of a major nuclear facility is an iterative process, which continues throughout the construction of the upgrades

  11. Design and implementation of a unified certification management system based on seismic business

    Science.gov (United States)

    Tang, Hongliang

    2018-04-01

    Many business software for seismic systems are based on web pages, users can simply open a browser and enter their IP address. However, how to achieve unified management and security management of many IP addresses, this paper introduces the design concept based on seismic business and builds a unified authentication management system using ASP technology.

  12. Design requirements, criteria and methods for seismic qualification of CANDU power plants

    International Nuclear Information System (INIS)

    Singh, N.; Duff, C.G.

    1979-10-01

    This report describes the requirements and criteria for the seismic design and qualification of systems and equipment in CANDU nuclear power plants. Acceptable methods and techniques for seismic qualification of CANDU nuclear power plants to mitigate the effects or the consequences of earthquakes are also described. (auth)

  13. PBMR phase 1 study: Seismic and structural design consideration - An overview of principles

    International Nuclear Information System (INIS)

    Wium, D.J.W.

    1997-01-01

    This paper briefly reviews the principles involved in the planning and design of the proposed facility to cater for seismic and structural loads. The conceptual layout is discussed, as well as the different load characteristics and scenarios. An outline is given of model used to estimate the seismic loads, whereafter the different analytical models are discussed. (author)

  14. Automatic seismic support design of piping system by an object oriented expert system

    International Nuclear Information System (INIS)

    Nakatogawa, T.; Takayama, Y.; Hayashi, Y.; Fukuda, T.; Yamamoto, Y.; Haruna, T.

    1990-01-01

    The seismic support design of piping systems of nuclear power plants requires many experienced engineers and plenty of man-hours, because the seismic design conditions are very severe, the bulk volume of the piping systems is hyge and the design procedures are very complicated. Therefore we have developed a piping seismic design expert system, which utilizes the piping design data base of a 3 dimensional CAD system and automatically determines the piping support locations and support styles. The data base of this system contains the maximum allowable seismic support span lengths for straight piping and the span length reduction factors for bends, branches, concentrated masses in the piping, and so forth. The system automatically produces the support design according to the design knowledge extracted and collected from expert design engineers, and using design information such as piping specifications which give diameters and thickness and piping geometric configurations. The automatic seismic support design provided by this expert system achieves in the reduction of design man-hours, improvement of design quality, verification of design result, optimization of support locations and prevention of input duplication. In the development of this system, we had to derive the design logic from expert design engineers and this could not be simply expressed descriptively. Also we had to make programs for different kinds of design knowledge. For these reasons we adopted the object oriented programming paradigm (Smalltalk-80) which is suitable for combining programs and carrying out the design work

  15. Seismic analysis of steam generator and parameter sensitivity studies

    International Nuclear Information System (INIS)

    Qian Hao; Xu Dinggen; Yang Ren'an; Liang Xingyun

    2013-01-01

    Background: The steam generator (SG) serves as the primary means for removing the heat generated within the reactor core and is part of the reactor coolant system (RCS) pressure boundary. Purpose: Seismic analysis in required for SG, whose seismic category is Cat. I. Methods: The analysis model of SG is created with moisture separator assembly and tube bundle assembly herein. The seismic analysis is performed with RCS pipe and Reactor Pressure Vessel (RPV). Results: The seismic stress results of SG are obtained. In addition, parameter sensitivities of seismic analysis results are studied, such as the effect of another SG, support, anti-vibration bars (AVBs), and so on. Our results show that seismic results are sensitive to support and AVBs setting. Conclusions: The guidance and comments on these parameters are summarized for equipment design and analysis, which should be focused on in future new type NPP SG's research and design. (authors)

  16. Recommended revisions to nuclear regulatory commission seismic design criteria

    International Nuclear Information System (INIS)

    Coats, D.W.

    1981-01-01

    Task Action Plan (TAP) A-40 was developed by consolidating specific technical assistance studies initiated to identify and quantify the conservatism inherent in the seismic design sequence of current NRC criteria. Task 10 of TAP A-40 provided a technical review of the results of the other nine engineering and seismological tasks in TAP A-40 and recommended changes to the existing NRC criteria based on this review. We used the team approach to accomplish the objectives of Task 10 in an efficient manner and to provide the best technical product possible within the limited time available. The team consisted of a core group of Lawrence Livermore National Laboratory personnel and selected consultants. The recommendations summarized in this paper were not based solely on the results of the tasks in TAP A-40 but went far beyond that data base to encompass all available and appropriate literature. Some recommendations are based on the expertise of core members and consultants that stem from unpublished data, research, and experience. Copies of the pertinent sections of the Standard Review Plan (SRP) and Regulatory Guides as well as the reports developed under TAP A-40 were provided to the participants. These reports, other available engineering literature, and the experience of the consultants and core group provided technical basis for the recommendations. (orig./HP)

  17. Temelin NPP commissioning experiences

    International Nuclear Information System (INIS)

    Hanus, V.

    2002-01-01

    The Building Permit for the Temelin NPP with four VVER units was issued in 1986, which is a long time ago. Since then, however, was taken a route that is very different from what anybody imagined. Described are the legislative and design changes and given is a current condition of the power plant

  18. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  19. Seismic design and analysis of nuclear fuel cycle facilities in France

    International Nuclear Information System (INIS)

    Sollogoub, P.

    2001-01-01

    Methodology for seismic design of nuclear fuel facilities and power plants in France is described. After the description of regulatory and normative texts for seismic design, different elements are examined: definition of ground motion, analysis methods, new trends, reevaluation and specificity of Fuel Cycle Facilities. R/D developments are explicated in each part. Their final objective are to better quantify the margins of each step which, in relation with safety analysis,lead to balanced design, analysis and retrofit rules. (author)

  20. Practice for the upgrading of Trino Vercellese NPP: Technical and economical aspects

    International Nuclear Information System (INIS)

    Giangrasso, M.; Maresca, G.; Pino, G.; Sano, T.

    1993-01-01

    In this report the experience gained in seismic re-evaluation of an old NPP (Trino Vercellese) is described. This PWR plant was not seismically designed. The main purpose of the upgrading, from the point of view of the Italian Directorate for Nuclear Safety - ENEA/DISP, was to have guaranteed the plant capability of achieving and maintaining a safe cold shutdown condition after a SSE seismic event. The main steps of the seismic review are discussed: definition of the new input motion; selection of structures, systems and components essential for a safe cold shutdown; definition of Codes and evaluation methods; seismic qualification of systems and components. Finally some modifications of a number of plant systems are described together with economical aspects. (author)

  1. Realistic design principles of nuclear power plants against earthquakes in the FRG - present stage of discussion of the new concept for a KTA-safety-standard concerning earthquake design

    International Nuclear Information System (INIS)

    Hintergraeber, M.; Wittmann, R.

    1985-01-01

    A new concept for the seismic design of npp was drafted in Germany. This new concept is thought to be a substitute of the existing safety standard KTA 2201.1 'Basic principles of the design of npp against seismic events' (issued 6/75). The aim of this presentation is to give a survey of the present stage of the relevant discussions within the regulatory committees. (orig.)

  2. Optimization Criteria In Design Of Seismic Isolated Building

    International Nuclear Information System (INIS)

    Clemente, Paolo; Buffarini, Giacomo

    2008-01-01

    Use of new anti-seismic techniques is certainly suitable for buildings of strategic importance and, in general, in the case of very high risk. For ordinary buildings, instead, the cost of base isolation system should be balanced by an equivalent saving in the structure. The comparison criteria have been first defined, then a large numerical investigation has been carried out to analyze the effectiveness and the economic suitability of seismic isolation in concrete buildings

  3. Design and development of digital seismic amplifier recorder

    Energy Technology Data Exchange (ETDEWEB)

    Samsidar, Siti Alaa; Afuar, Waldy; Handayani, Gunawan, E-mail: gunawanhandayani@gmail.com [Department of Physics, ITB (Indonesia)

    2015-04-16

    A digital seismic recording is a recording technique of seismic data in digital systems. This method is more convenient because it is more accurate than other methods of seismic recorders. To improve the quality of the results of seismic measurements, the signal needs to be amplified to obtain better subsurface images. The purpose of this study is to improve the accuracy of measurement by amplifying the input signal. We use seismic sensors/geophones with a frequency of 4.5 Hz. The signal is amplified by means of 12 units of non-inverting amplifier. The non-inverting amplifier using IC 741 with the resistor values 1KΩ and 1MΩ. The amplification results were 1,000 times. The results of signal amplification converted into digital by using the Analog Digital Converter (ADC). Quantitative analysis in this study was performed using the software Lab VIEW 8.6. The Lab VIEW 8.6 program was used to control the ADC. The results of qualitative analysis showed that the seismic conditioning can produce a large output, so that the data obtained is better than conventional data. This application can be used for geophysical methods that have low input voltage such as microtremor application.

  4. Seismic design ampersand analysis considerations for high level nuclear waste repositories

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1993-01-01

    A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid open-quotes deterministicclose quotes and open-quotes probabilisticclose quotes concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

  5. Seismic design and analysis considerations for high level nuclear waste repositories

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1993-01-01

    A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid ''deterministic'' and ''probabilistic'' concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

  6. Design approach of seismic interface for cryoline with Tokamak building for ITER

    International Nuclear Information System (INIS)

    Badgujar, S.; Sarkar, B.; Vaghela, H.; Shah, N.; Naik, H.B.

    2012-01-01

    ITER Tokamak building is designed with seismic isolation pads to protect the Tokamak components from seismic events. Two main cryolines, designated as cryolines between buildings (Mg and CP), runs from interconnection box in cryoplant building to the Tokamak building. The lines outside Tokamak building are supported by seismically non-isolated supports. The cryoline design at the interface between seismically isolated and non-isolated support systems needs to be studied to fulfill the functional requirements. One of the options for interface, universal expansion joint has been modeled in CATIA with actual thickness of each ply and inter-ply distance, analyzed in ANSYS using contact definition, as a part of the preliminary study. The bellows have been checked by design calculation as per EJMA standard for the specified movements. The paper will present approach for conceptual design of interface, problem definition and boundary conditions, methodology for analysis and preliminary results of stress pattern for expansion joints. (author)

  7. Experience for plant monitoring design in Italian BWR NPP and future trends in man-machine interface

    International Nuclear Information System (INIS)

    Maestri, F.; Sepielli, M.

    1987-01-01

    TMI accidental sequence and daily-gained operating experience on italian and abroad NPPs have affected in depth the approach to the design of information presentation to the Control Room staff. It has been cleared that most problems in plant operation arise from a poor and inadequate information system. The main lacks have been identified in the Control Room lay-out and information organization. This has pushed designers both to improve the Control Room environment and to better exploit the computer data processing and data presentation capabilities. The paper deals with the basic criteria for the design and the design review of a computerized system to be inserted in a hybrid Control Room in Italian 981 Mwe BWR-6 NPP, where the concepts outlined above were taken-up from the very beginning. The Control Room keeps conventional instrumentation arranged in a human-factor lay-out, according to post-TMI requirements, and adds a powerful computer-based information system for advanced alarm presentation and plant supervision during both normal and emergency conditions with high data reliability. Colour videounits and operating panels are functionally integrated to create powerful operator work-stations. Emphasis is mostly given on the revision work for video-unit displays and Man-System Communication carried out in cooperation with Halden Reactor Project human factor and plant operation experts. The work peculiarity has been a strong care on the integration between conventional and computerized information presentation, with particular regard to common information and code consistency. (author)

  8. Regulatory review of NPP Krsko Periodic Safety Review

    International Nuclear Information System (INIS)

    Lovincic, D.; Muehleisen, A.; Persic, A.

    2004-01-01

    At the request of the Slovenian Nuclear Safety Administration (SNSA), Krsko NPP prepared a Periodic Safety Review (PSR) program in January 2001. This is the first PSR of NPP Krsko, the only nuclear power plant in Slovenia. The program was reviewed by the IAEA mission in May 2001 and approved by SNSA in July 2001. The program is made in accordance with the IAEA Safety Guide 'Periodic Safety Review of Operational Nuclear Power Plants' No. 50-SG-012 and with European practice. It contains a systematic review of operation of the NPP Krsko, including the review of the changes as a result of the modernization of the facility. The main tasks of PSR are review of plant status for each safety factor, development of aging and life cycle management program, review of seismic design and PSHA analysis and update of regulatory compliance program. The prioritization process of findings and action plan are also important tasks of PSR. The basic safety factors of the PSR review are: Operational Experience, Safety Assessment and Analyses, Equipment Qualification and Ageing Management, Safety Culture, Emergency Planing, Environmental Impact and Radioactive Waste, Compliance with license requirements and Prioritization. It had been agreed that SNSA will have reviewed all PSR reports generated during the PSR process. At the end of 2003 the PSR Summary Report with selected recommendations for action plan was completed and delivered to SNSA for review. The paper presents regulatory review of NPP Krsko PSR with emphasis on the evaluation of the PSR issues ranking process. (author)

  9. Overcoming barriers to high performance seismic design using lessons learned from the green building industry

    Science.gov (United States)

    Glezil, Dorothy

    NEHRP's Provisions today currently governing conventional seismic resistant design. These provisions, though they ensure the life-safety of building occupants, extensive damage and economic losses may still occur in the structures. This minimum performance can be enhanced using the Performance-Based Earthquake Engineering methodology and passive control systems like base isolation and energy dissipation systems. Even though these technologies and the PBEE methodology are effective reducing economic losses and fatalities during earthquakes, getting them implemented into seismic resistant design has been challenging. One of the many barriers to their implementation has been their upfront costs. The green building community has faced some of the same challenges that the high performance seismic design community currently faces. The goal of this thesis is to draw on the success of the green building industry to provide recommendations that may be used overcome the barriers that high performance seismic design (HPSD) is currently facing.

  10. Verifying seismic design of nuclear reactors by testing. Volume 2: appendix, theoretical discussions

    International Nuclear Information System (INIS)

    1979-01-01

    Theoretical discussions on seismic design testing are presented under the following appendix headings: system functions, pulse optimization program, system identification, and motion response calculations from inertance measurements of a nuclear power plant

  11. Outline of the seismic design guideline of an FBR - a tentative draft

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Ohtsubo, Hideomi; Nakamura, Hideharu; Matsuura, Shinichi; Hagiwara, Yutaka; Yuhara, Tetsuo; Hirayama, Hiroshi; Kokubo, Kunio; Ooka, Yuji.

    1993-01-01

    Central Research Institute of Electric Power Industry (Japan) is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings. (orig.)

  12. The proposed human factors engineering program plan for man-machine interface system design of the next generation NPP in Korea

    International Nuclear Information System (INIS)

    Oh, I.S.; Lee, H.C.; Seo, S.M.; Cheon, S.W.; Park, K.O.; Lee, J.W.; Sim, B.S.

    1994-01-01

    Human factors application to nuclear power plant (NPP) design, especially, to man-machine interface system (MMIS) design becomes an important issue among the licensing requirements. Recently, the nuclear regulatory bodies require the evidence of systematic human factors application to the MMIS design. Human Factors Engineering Program Plan (HFEPP), as a basis and central one among the human factors application by the MMIS designers. This paper describes the framework of HFEPP for the MMIS design of next generation NPP (NG-NPP) in Korea. This framework provides an integral plan and some bases of the systematic application of human factors to the MMIS design, and consists of purpose and scope, codes and standards, human factors organization, human factors tasks, engineering control methodology, human factors documentations, and milestones. The proposed HFEPP is a top level document to define and describe human factors tasks, based on each step of MMIS design process, in view point of how, what, when and by whom to be performed. (author). 11 refs, 1 fig

  13. Geological-Hydrological Site Evaluation for NPP Planning

    Energy Technology Data Exchange (ETDEWEB)

    Faust, Brigitte; Mini, Paolo [Nordostsschweizerische Kraftwerke AG NOK, Parkstrasse 23, 5401 Baden (Switzerland)

    2008-07-01

    NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

  14. Geological-Hydrological Site Evaluation for NPP Planning

    International Nuclear Information System (INIS)

    Faust, Brigitte; Mini, Paolo

    2008-01-01

    NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

  15. Seismic design criteria for the system 80+ advanced light water reactor

    International Nuclear Information System (INIS)

    Manrique, M.A.; Dermitzakis, S.N.; Gerdes, L.D.; Kennedy, R.P.; Idriss, I.M.; Cassidy, J.R.

    1991-01-01

    This paper presents the development of seismic design criteria in support of design certification by the Nuclear Regulatory Commission (NRC) of the ABB-Combustion Engineering's System 80+ Standard Design. The design certification effort is sponsored by the US Department of Energy (DOE). The development of the design criteria included: (a) development of the seismic control motion, (b) development of generic soil profiles for anticipated sites, (c) generation of in-structure response spectra and design loads for structures and equipment through soil-structure interaction (SSI) analyses, and (d) acceptance criteria for future construction sites

  16. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy

  17. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy.

  18. Seismic Response Analysis and Design of Structure with Base Isolation

    International Nuclear Information System (INIS)

    Rosko, Peter

    2010-01-01

    The paper reports the study on seismic response and energy distribution of a multi-story civil structure. The nonlinear analysis used the 2003 Bam earthquake acceleration record as the excitation input to the structural model. The displacement response was analyzed in time domain and in frequency domain. The displacement and its derivatives result energy components. The energy distribution in each story provides useful information for the structural upgrade with help of added devices. The objective is the structural displacement response minimization. The application of the structural seismic response research is presented in base-isolation example.

  19. Role of PRA in new NPP projects

    International Nuclear Information System (INIS)

    Julin, A.; Sandberg, J.; Virolainen, R.

    2012-01-01

    In Finland, a plant specific, Level 1 and 2 Probabilistic Risk Analysis (PRA) is required as a prerequisite for issuing the construction license and operating license. The use of PRA in various applications and the main insights are presented. These applications include e.g. PRA support to the design of SSCs (Systems, Structures and Components), definition of pre-service and in-service inspection programs, evaluation of the safety classification of SSCs, development of procedures, training and in definition of risk informed technical specifications, periodic testing and on-line preventive maintenance programs. In addition, PRA shall be used to assess the adequacy and coverage of the phase and system commissioning programs. Also the potential risks related to commissioning tests during nuclear test phase, shall be assessed with the help of PRA. In OL3 project, risk informed approach has been applied on a large scale for the first time in the design, construction and commissioning of a new NPP unit. Pre-nuclear commissioning tests have started at OL3 site and the plant is foreseen to begin commercial operation in 2013. Decisions have been made to launch new NPP projects. Teollisuuden Voima Oyj (TVO) is planning to build a new unit (OL4) at Olkiluoto site and a new utility, Fennovoima, is planning to build one unit at one of two alternative green field sites in Northern parts of Finland. Insights from PRAs of operating NPPs have been used in the evaluation of possible new sites to ensure that the site specific concerns and environmental conditions are adequately taken into account in the design of SSCs. Although the seismic activity at the Olkiluoto site is low, a comprehensive seismic risk analysis is being conducted. Its results support the review of the deterministic seismic design. For new sites, a probabilistic seismic hazard analysis has been carried out for the determination of the design earthquake. Experiences from OL3 licensing have been utilized in the

  20. Yield Frequency Spectra and seismic design of code-compatible RC structures: an illustrative example

    DEFF Research Database (Denmark)

    Katsanos, Evangelos; Vamvatsikos, Dimitrios

    2017-01-01

    with given yield displacement and capacity curve shape. For the 8-story case study building, deformation checking is the governing limit state. A conventional code-based design was performed using seismic intensities tied to the desired MAF for safety checking. Then, the YFS-based approach was employed......The seismic design of an 8-story reinforced concrete space frame building is undertaken using a Yield Frequency Spectra (YFS) performance-based approach. YFS offer a visual representation of the entire range of a system’s performance in terms of the mean annual frequency (MAF) of exceeding...... to redesign the resulting structure working backwards from the desired MAF of response (rather than intensity) to estimate an appropriate value of seismic intensity for use within a typical engineering design process. For this high-seismicity and high-importance midrise building, a stiffer system with higher...

  1. Regulatory requirements for radiation safety in the design of a new Finish NPP

    Energy Technology Data Exchange (ETDEWEB)

    Alm-Lytz, Kirsi; Vilkamo, Olli [Radiation and Nuclear Safety Authority, STUK, PO Box 14, Laippatie 4, 00881 Helsinki (Finland)

    2004-07-01

    There are two operating nuclear power plants in Finland, two BWR units at Olkiluoto site and two PWR units at Loviisa site. These reactors were commissioned between 1977 and 1981. The total electricity capacity in Finland is about 15 GW. In 2003, nuclear power plants generated one fourth of Finland's electricity. Despite of the diversity of the electricity generation methods, Finland is highly dependent on imported energy. Electricity consumption is estimated to increase and the demand for extra capacity has been estimated at about 2500-3000 MW by 2010. It should also be taken into account that a considerable proportion of the production capacity constructed in the 1970's must be replaced with production capacity of new power plants in the near future. In practice, the climate politics commitments made by Finland exclude coal power. Therefore, the capacity can be increased significantly only by natural gas, nuclear power and biofuels. The paper presents the following issues: Licensing a new nuclear power plant in Finland; FIN5 Project at STUK; Work planning and a tool for requirement management; Radiation safety related YVL guides; Collective dose target; On-site habitability during accident situation. Habitability was evaluated on the basis of the calculated dose rate levels, the occupancy times and the dose limits. Radiation hazard was classified into three parts, i.e., possible direct radiation from the containment, air contamination and systems carrying radioactive air or water. The results showed that direct radiation from the containment is generally adequately shielded but penetrations and hatches have to be separately analysed and the radiation dose levels near them are usually rather high. Skyshine radiation from the reactor containment is a special feature at the Loviisa NPP and the nearby area outside the buildings might have very limited access for the first hours after the accident. The skyshine effect is not usually relevant hazard in

  2. Consideration on the relation between dynamic seismic motion and static seismic coefficient for the earthquake proof design of slope around nuclear power plant

    International Nuclear Information System (INIS)

    Ito, Hiroshi; Kitahara, Yoshihiro; Hirata, Kazuta

    1986-01-01

    When the large cutting slopes are constructed closed to around nuclear power plants, it is important to evaluate the stability of the slopes during the strong earthquake. In the evaluation, it may be useful to clarify relationship between the static seismic coefficient and dynamic seismic force corresponded to the basic seismic motion which is specified for designing the nuclear power facilities. To investigate this relation some numerical analyses are conducted in this paper. As the results, it is found that dynamic forces considering the amplified responses of the slopes subjected to the basic seismic motion with a peak acceleration of 500 gals at the toe of the slopes, are approximately equal to static seismic force which generates in the slopes when the seismic coefficients of k = 0.3 is applied. (author)

  3. Investigation of optimal seismic design methodology for piping systems supported by elasto-plastic dampers. Part. 2. Applicability for seismic waves with various frequency characteristics

    International Nuclear Information System (INIS)

    Ito, Tomohiro; Michiue, Masashi; Fujita, Katsuhisa

    2010-01-01

    In this study, the applicability of a previously developed optimal seismic design methodology, which can consider the structural integrity of not only piping systems but also elasto-plastic supporting devices, is studied for seismic waves with various frequency characteristics. This methodology employs a genetic algorithm and can search the optimal conditions such as the supporting location and the capacity and stiffness of the supporting devices. Here, a lead extrusion damper is treated as a typical elasto-plastic damper. Numerical simulations are performed using a simple piping system model. As a result, it is shown that the proposed optimal seismic design methodology is applicable to the seismic design of piping systems subjected to seismic waves with various frequency characteristics. The mechanism of optimization is also clarified. (author)

  4. The Ductile Design Concept for Seismic Actions in Miscellaneous Design Codes

    Directory of Open Access Journals (Sweden)

    M. Budescu

    2009-01-01

    Full Text Available The concept of ductility estimates the capacity of the structural system and its components to deform prior to collapse, without a substantial loss of strength, but with an important energy amount dissipated. Consistent with the „Applied Technology Council” (ATC-34, from 1995, it was agreed that the reduction seismic response factor to decrease the design force. The purpose of this factor is to transpose the nonlinear behaviour of the structure and the energy dissipation capacity in a simplified form that can be used in the design stage. Depending on the particular structural model and the design standard the used values are different. The paper presents the characteristics of the ductility concept for the structural system. Along with this the general way of computing the reserve factor with the necessary explanations for the parameters that determine the behaviour factor are described. The purpose of this paper is to make a comparison between different international norms for the values and the distribution of the behaviour factor. The norms from the following countries are taken into consideration: the United States of America, New Zealand, Japan, Romania and the European general seismic code.

  5. NSR&D Program Fiscal Year (FY) 2015 Call for Proposals Mitigation of Seismic Risk at Nuclear Facilities using Seismic Isolation

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Seismic isolation (SI) has the potential to drastically reduce seismic response of structures, systems, or components (SSCs) and therefore the risk associated with large seismic events (large seismic event could be defined as the design basis earthquake (DBE) and/or the beyond design basis earthquake (BDBE) depending on the site location). This would correspond to a potential increase in nuclear safety by minimizing the structural response and thus minimizing the risk of material release during large seismic events that have uncertainty associated with their magnitude and frequency. The national consensus standard America Society of Civil Engineers (ASCE) Standard 4, Seismic Analysis of Safety Related Nuclear Structures recently incorporated language and commentary for seismically isolating a large light water reactor or similar large nuclear structure. Some potential benefits of SI are: 1) substantially decoupling the SSC from the earthquake hazard thus decreasing risk of material release during large earthquakes, 2) cost savings for the facility and/or equipment, and 3) applicability to both nuclear (current and next generation) and high hazard non-nuclear facilities. Issue: To date no one has evaluated how the benefit of seismic risk reduction reduces cost to construct a nuclear facility. Objective: Use seismic probabilistic risk assessment (SPRA) to evaluate the reduction in seismic risk and estimate potential cost savings of seismic isolation of a generic nuclear facility. This project would leverage ongoing Idaho National Laboratory (INL) activities that are developing advanced (SPRA) methods using Nonlinear Soil-Structure Interaction (NLSSI) analysis. Technical Approach: The proposed study is intended to obtain an estimate on the reduction in seismic risk and construction cost that might be achieved by seismically isolating a nuclear facility. The nuclear facility is a representative pressurized water reactor building nuclear power plant (NPP) structure

  6. SEISMIC DESIGN OF TWO STOREY REINFORCED CONCRETE BUILDING IN MALAYSIA WITH LOW CLASS DUCTILITY

    OpenAIRE

    MOHD IRWAN ADIYANTO; TAKSIAH A. MAJID

    2014-01-01

    Since Malaysia is not located in active seismic fault zones, majority of buildings in Malaysia had been designed according to BS8110, which not specify any seismic provision. After experienced several tremors originating from neighbouring countries especially from Sumatra, Indonesia, the Malaysian start to ask questions on integrity of existing structures in Malaysia to withstand the earthquake load. The question also arises regarding the economical effect in term of cost of construction if s...

  7. SEISMIC DESIGN OF TWO STOREY REINFORCED CONCRETE BUILDING IN MALAYSIA WITH LOW CLASS DUCTILITY

    Directory of Open Access Journals (Sweden)

    MOHD IRWAN ADIYANTO

    2014-02-01

    Full Text Available Since Malaysia is not located in active seismic fault zones, majority of buildings in Malaysia had been designed according to BS8110, which not specify any seismic provision. After experienced several tremors originating from neighbouring countries especially from Sumatra, Indonesia, the Malaysian start to ask questions on integrity of existing structures in Malaysia to withstand the earthquake load. The question also arises regarding the economical effect in term of cost of construction if seismic design has to be implemented in Malaysian construction industry. If the cost is increasing, how much the increment and is it affordable? This paper investigated the difference of steel reinforcement and concrete volume required when seismic provision is considered in reinforced concrete design of 2 storey general office building. The regular office building which designed based on BS8110 had been redesigned according to Eurocode 2 with various level of reference peak ground acceleration, agR reflecting Malaysian seismic hazard for ductility class low. Then, the all frames had been evaluated using a total of 800 nonlinear time history analyses considering single and repeated earthquakes to simulate the real earthquake event. It is observed that the level of reference peak ground acceleration, agR and behaviour factor, q strongly influence the increment of total cost. For 2 storey RC buildings built on Soil Type D with seismic consideration, the total cost of material is expected to increase around 6 to 270%, depend on seismic region. In term of seismic performance, the repeated earthquake tends to cause increasing in interstorey drift ratio around 8 to 29% higher compared to single earthquake.

  8. Considerations for developing seismic design criteria for nuclear waste storage repositories

    International Nuclear Information System (INIS)

    Owen, G.N.; Yanev, P.I.; Scholl, R.E.

    1980-04-01

    The function of seismic design criteria is to reduce the potential for hazards that may arise during various stages of the repository life. During the operational phase, the major concern is with the possible effects of earthquakes on surface facilities, underground facilities, and equipment. During the decommissioned phase, the major concern is with the potential effects of earthquakes on the geologic formation, which may result in a reduction in isolation capacity. Existing standards and guides or criteria used for the static and seismic design of licensed nuclear facilities were reviewed and evaluated for their applicability to repository design. This report is directed mainly toward the development of seismic design criteria for the underground structures of repositories. An initial step in the development of seismic design criteria for the underground structures of repositories is the development of performance criteria, or minimum standards of acceptable behavior. A number of possible damage modes are identified for the operating phase of the repository; however, no damage modes are foreseen that would perturb the long-term function of the repository, except for the possibility of increased permeability within the rock mass. Subsequent steps in formulating acceptable seismic design criteria for the underground structures involve the quantification of the design process. The report discusses the necessity of specifying the form of ground motion that would be needed for seismic analysis and the procedures that may be used for making ground motion predictions. Further discussions outline what is needed for analysis, including rock properties, failure criteria, modeling techniques, seismic hardening criteria for the host rock mass, and probabilistic considerations

  9. Use of MAAP code for identification of key plant vulnerabilities for the beyond design accidents and their mitigation at NPP Krsko

    International Nuclear Information System (INIS)

    Krajnc, B.

    1995-01-01

    NPP Krsko performed according to GL 88-20, Supplement 1-4 and RUJV requirement the Individual Plant Examination analyses. For the required deterministic analyses the MAAP 3.0B code was used. It was proven that such severe accident analysis can be used for evaluation of the overall level of safety improvement that can be gained with the different modifications and alternate design. In this paper one such important outcomes from these analyses will be presented. (author)

  10. Computer-aided design system for a complex of problems on calculation and analysis of engineering and economical indexes of NPP power units

    International Nuclear Information System (INIS)

    Stepanov, V.I.; Koryagin, A.V.; Ruzankov, V.N.

    1988-01-01

    Computer-aided design system for a complex of problems concerning calculation and analysis of engineering and economical indices of NPP power units is described. In the system there are means for automated preparation and debugging of data base software complex, which realizes th plotted algorithm in the power unit control system. Besides, in the system there are devices for automated preparation and registration of technical documentation

  11. Seismic design assessment by experimental methods. Notes from the workshop. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The workshop intended to provide training on the application of experimental techniques (mainly laboratory testing) as support to the seismic design of structures, equipment and components for nuclear power plants. The focus was on the activities planned by Nuclear Power Institute of China (NPIC) in the near future, and most of the lectures provided by the attendees, dealing with these national activities, were the basis for the discussion with the IAEA experts. Special modules were identified for the workshop, dealing with: Numerical models: detailing and comparison techniques; On site testing of structures and equipment; Special problems: Leak before Break (LBB), thermal effects, combination of seismic with other loads; General seismic behavior and design criteria for fuel assembly and core structures; Seismic qualification methodologies for reactor core, mechanical components, I and C and piping; Balancing analysis and test in seismic qualification; Design of mock-up: selection of seismic input, detailing, scaling and similitudes, selection of sensors and their location; Test planning and conduct, basic documents and specifications; Quality assurance and technical procedures in laboratory testing; Data processing techniques and interface with the numerical models. The material used for presentations by the lecturers and by the national attendees is collected in this volume together with some background literature provided by the experts with up to date references and procedures. A special chapter is added to these proceedings with the content of the discussion, for future reference and as a complement to the lectures content, more oriented to the specific, immediate needs of the attendees.

  12. Seismic design assessment by experimental methods. Notes from the workshop. Working material

    International Nuclear Information System (INIS)

    2001-01-01

    The workshop intended to provide training on the application of experimental techniques (mainly laboratory testing) as support to the seismic design of structures, equipment and components for nuclear power plants. The focus was on the activities planned by Nuclear Power Institute of China (NPIC) in the near future, and most of the lectures provided by the attendees, dealing with these national activities, were the basis for the discussion with the IAEA experts. Special modules were identified for the workshop, dealing with: Numerical models: detailing and comparison techniques; On site testing of structures and equipment; Special problems: Leak before Break (LBB), thermal effects, combination of seismic with other loads; General seismic behavior and design criteria for fuel assembly and core structures; Seismic qualification methodologies for reactor core, mechanical components, I and C and piping; Balancing analysis and test in seismic qualification; Design of mock-up: selection of seismic input, detailing, scaling and similitudes, selection of sensors and their location; Test planning and conduct, basic documents and specifications; Quality assurance and technical procedures in laboratory testing; Data processing techniques and interface with the numerical models. The material used for presentations by the lecturers and by the national attendees is collected in this volume together with some background literature provided by the experts with up to date references and procedures. A special chapter is added to these proceedings with the content of the discussion, for future reference and as a complement to the lectures content, more oriented to the specific, immediate needs of the attendees

  13. Development of Cold-Formed Steel Seismic Design Recommendations

    Science.gov (United States)

    2015-08-01

    top beam attached to the hydraulic ram. This tube is far more flexible in bending than the beam/floor slab in the field, and therefore will not... torsional and out-of-plane response should not be significant, and the rocking response can be accounted for in refined Drain 2DX analysis when...209 11.9.3 Vertical distribution of lateral seismic forces ........................................................ 210 11.9.4 Torsion

  14. Optimizing the NPP refurbishment environmental qualification design process for an in-service installation

    International Nuclear Information System (INIS)

    MacBeth, M.J.; Hemmings, R.L.

    2002-01-01

    This paper describes the Environmental Qualification (EQ) modification design work required to upgrade the Reactor Building (RB) airlocks at the Pickering Nuclear Generating Station B Facility. The RB airlocks provide a containment boundary function and are designed to prevent a breach of containment from occurring for all analysed station conditions. Recent, more stringent, Canadian Nuclear Regulatory actions have imposed EQ requirements for the RB airlocks in Canadian Nuclear Generating Stations. The airlocks are required to function under the worst-case design basis accident (DBA) conditions for the assigned mission duration and the design must demonstrate that a spurious door opening cannot be initiated by any accident conditions. This paper reviews key project design activities while providing detailed insights to the potential solution or elimination of some problematic aspects of these types of design activities. General recommendations for optimal technical management of such project implementation issues are presented. (author)

  15. Design of the connection pieces between concrete and valves of the cooling water system in the Angra I NPP turbine building

    International Nuclear Information System (INIS)

    Diaz, B.E.; Carvalho, L.J. de

    1988-01-01

    This work describes the design characteristics of the Transition Pieces between concrete galleries and valves of the Cooling Water System of the Turbine Building of the Angra NPP-Unit I. Design details concerning the structure and procedures for the structural analysis are presented. It is emphasized that the usual simplified design rules for the flange and bolts can not be used in the case of non existent polar symmetry for the structure and applied loads. A more sophisticated design based on finite elements models is required in these cases. (author) [pt

  16. Accident beyond the design basis management with the coolant loss at the NPP with WWER

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

    2010-01-01

    The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

  17. External hazards analysis approach to level 1 PSA of Mochovce NPP - Slovakia

    International Nuclear Information System (INIS)

    Stojka, Tibor

    2000-01-01

    Analyses of external events had been first time performed at the design stage of the Mochovce NPP showing sufficiently low contribution of external hazards to core damage frequency. But, based on IAEA document 'Safety problems of WWER-440/213 NPPs and the categorization' (IAEA-EBP-WWER-03, 1996), the need of new reassessment arose due to discrepancy of some origin recommendations in compare with present IAEA ones. Mochovce NPP Nuclear Safety Improvements Program elaborated at the same time included the IAEA recommendations and following improvements were proposed to perform in context of external events. 1. Seismic project and new locality seismic evaluation This safety improvement includes also some 'on site' technical improvements in seismic stability of structures and equipment. 2. Unit specific analyses of extreme meteorologic conditions. This safety improvement focuses on impact of feasible extreme conditions on NPP systems caused by rain, snow and hail storms, frost, winds, low and high temperatures. 3. Analyses of external hazards caused by humans. In this safety improvement were specified: feasible sources of explosions; analyses of hydrogen, gas and propane-calor gas depots; air crash risk. The results of these implemented safety improvements were considered in the PSA study. The External hazards analysis is also part of Level 1 PSA Mochovce NPP performed by PSA Department of VUJE Trnava Inc., Engineering, Design and Research Organization, Slovakia. Some partial analyses are performed in cooperation with following companies DS and S - SAIC, USA and Geophysical Institute Academy of Science, Slovakia Relko, Slovakia. Basic documents are: NUREG/CR-2300 'PRA Procedures Guide - A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants' and IAEA SS No. 50-P-7 'Treatment of External Hazards in PSA for NPPs. The external hazards analysis consists of following parts: 1. Geography and plant locality; 2. Nearby industry; 3. Extreme

  18. Structural seismic upgrading of NPPs in Czech and Slovak republics

    Energy Technology Data Exchange (ETDEWEB)

    David, M [DAVID Consulting, Engineering and Design Office, Prague (Czech Republic)

    1997-03-01

    Several Nuclear Power Plants of the VVER type has been constructed during the past years in former Czechoslovak Republic. Some of them has been already put in operation and some of them are under construction. Nuclear Power Plants V1(2 units of VVER 440/230), V2(2 units of VVER 440/213) in Slovak and NPP Dukovany (4 units of VVER 440/213) in Czech republic are in operation. NPP Mochovce (4 units of VVER 440/213) in Slovak and NPP Temelin (4 units reduced now to 2 units VVER 1000) have been already almost completed, but still under construction. All above cited NPPs have not been either explicitly designed against earthquake or the design against earthquake or its input data must be upgraded to be compatible with present requirements. The upgrading of seismic input as well the seismic upgrading of all structures and technological equipments for so many NPPs has involved a lot of comprehensive work in Czech as well as in Slovak republics. The upgrading cannot be completed in a short time and as a rule the seismic upgrading has been usually performed in several steps, beginning with the most important arrangements against seismic hazard. The basic principles and requirements for seismic upgrading has been defined in accordance with the international and particularly with the IAEA recommendations. About the requirements for seismic upgrading of NPPs in Czech and Slovak republics will be reported in other contribution. This contribution is dealing with the problems of seismic upgrading of NNPs civil engineering structures. The aim of this contribution is to point out some specific problems connected firstly with very complicated concept of Versa structures and secondly with the difficult task to increase the structural capacity to the required seismic level. (J.P.N.)

  19. Structural seismic upgrading of NPPs in Czech and Slovak republics

    International Nuclear Information System (INIS)

    David, M.

    1997-01-01

    Several Nuclear Power Plants of the VVER type has been constructed during the past years in former Czechoslovak Republic. Some of them has been already put in operation and some of them are under construction. Nuclear Power Plants V1(2 units of VVER 440/230), V2(2 units of VVER 440/213) in Slovak and NPP Dukovany (4 units of VVER 440/213) in Czech republic are in operation. NPP Mochovce (4 units of VVER 440/213) in Slovak and NPP Temelin (4 units reduced now to 2 units VVER 1000) have been already almost completed, but still under construction. All above cited NPPs have not been either explicitly designed against earthquake or the design against earthquake or its input data must be upgraded to be compatible with present requirements. The upgrading of seismic input as well the seismic upgrading of all structures and technological equipments for so many NPPs has involved a lot of comprehensive work in Czech as well as in Slovak republics. The upgrading cannot be completed in a short time and as a rule the seismic upgrading has been usually performed in several steps, beginning with the most important arrangements against seismic hazard. The basic principles and requirements for seismic upgrading has been defined in accordance with the international and particularly with the IAEA recommendations. About the requirements for seismic upgrading of NPPs in Czech and Slovak republics will be reported in other contribution. This contribution is dealing with the problems of seismic upgrading of NNPs civil engineering structures. The aim of this contribution is to point out some specific problems connected firstly with very complicated concept of Versa structures and secondly with the difficult task to increase the structural capacity to the required seismic level. (J.P.N.)

  20. Original seismic design data and application of SMA and GIP methodologies. V. 1

    International Nuclear Information System (INIS)

    Masopust, R.

    1995-01-01

    The major focus of the IAEA sponsored Benchmark study for seismic analysis of WWER type NPPs is to develop the procedures which should be recommended to assess and enhance the seismic capacity of existing NPPs. The main issues are; identification of the most critical systems, structures and components necessary for safe shutdown; evaluation of as built conditions by collecting the data as originally used codes and standards, design drawings and construction specifications; realistic assessment of seismic response of plant building structures, distribution systems and passive equipment; functional qualification of active mechanical and electrical components through the use seismic experience or test-based data. The main aim of this report is to present the contribution to the task 'Safe shutdown system identification and classification'; to report on the task 'Standards, Criteria - Comparative study'; to present some special considerations coherent to these tasks

  1. The 1995 forum on appropriate criteria and methods for seismic design of nuclear piping

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1996-01-01

    A record of the 1995 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping is provided. The focus of the forum was the earthquake experience data base and whether the data base demonstrates that seismic inertia loads will not cause failure in ductile piping systems. This was a follow-up to the 1994 Forum when the use of earthquake experience data, including the recent Northridge earthquake, to justify a design-by-rule method was explored. Two possible topics for the next forum were identified--inspection after an earthquake and design for safe-shutdown earthquake only

  2. The Seismic Fragility Evaluation of an Offsite Transformer according to Aging Effects

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Choi, In Kil

    2008-01-01

    A seismic fragility analysis was performed, especially for an aged electric power transmission system, in this study. A real electric transformer system for Korean Nuclear Power Plants was selected for the seismic fragility evaluation. In the case of a seismic fragility analysis we should use design material properties and conditions. However material properties and environmental conditions of most structures and equipment are changed according to a lapse of time. Aging conditions greatly affect the integrity of the structures and equipment at NPP sites, but it is very difficult to estimate them qualitatively. Integrity of an anchor bolt system was considered with the aging conditions for an electric transformer system. At first, a seismic fragility analysis was performed for a fine condition for an electric transformer system. After that, a seismic fragility analysis according to the fastener of an anchor bolt system was conducted. This study showed that a looser anchor bolt creates seismic responses and seismic fragility changes of more 10%

  3. Evaluation of seismic behavior of soils under nuclear containment structures via dynamic centrifuge test

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Jeong Gon, E-mail: jgha87@kaist.ac.kr; Kim, Dong-Soo, E-mail: dskim@kaist.ac.kr

    2014-10-01

    Highlights: • A series of dynamic centrifuge tests were performed for NPP structure to investigate the soil–foundation-structure interaction with various soil conditions from loose sand to weathered rock. • SFSI phenomena for NPP structure were observed directly using experimental method. • Effect of the soil stiffness and nonlinear characteristics on SFSI was estimated. • There are comparisons of the control motions for seismic design of a NPP structure. • Subsoil condition, earthquake intensity and control motion affected to seismic load. - Abstract: To evaluate the earthquake loads for the seismic design of a nuclear containment structure, it is necessary to consider the soil–foundation-structure interaction (SFSI) due to their interdependent behavior. Especially, understanding the effects of soil stiffness under the structure and the location of control motion to SFSI are very important. Motivated by these requirements, a series of dynamic centrifuge tests were performed with various soil conditions from loose sand to weathered rock (WR), as well as different seismic intensities for the bedrock motion. The different amplification characteristics in peak-accelerations profile and effects of soil-nonlinearity in response spectrum were observed. The dynamic behaviors were compared between surface of free-field and foundation of the structure for the evaluation of the control motion for seismic design. It was found that dynamic centrifuge test has potentials to estimate the seismic load considering SFSI.

  4. Evaluation of seismic behavior of soils under nuclear containment structures via dynamic centrifuge test

    International Nuclear Information System (INIS)

    Ha, Jeong Gon; Kim, Dong-Soo

    2014-01-01

    Highlights: • A series of dynamic centrifuge tests were performed for NPP structure to investigate the soil–foundation-structure interaction with various soil conditions from loose sand to weathered rock. • SFSI phenomena for NPP structure were observed directly using experimental method. • Effect of the soil stiffness and nonlinear characteristics on SFSI was estimated. • There are comparisons of the control motions for seismic design of a NPP structure. • Subsoil condition, earthquake intensity and control motion affected to seismic load. - Abstract: To evaluate the earthquake loads for the seismic design of a nuclear containment structure, it is necessary to consider the soil–foundation-structure interaction (SFSI) due to their interdependent behavior. Especially, understanding the effects of soil stiffness under the structure and the location of control motion to SFSI are very important. Motivated by these requirements, a series of dynamic centrifuge tests were performed with various soil conditions from loose sand to weathered rock (WR), as well as different seismic intensities for the bedrock motion. The different amplification characteristics in peak-accelerations profile and effects of soil-nonlinearity in response spectrum were observed. The dynamic behaviors were compared between surface of free-field and foundation of the structure for the evaluation of the control motion for seismic design. It was found that dynamic centrifuge test has potentials to estimate the seismic load considering SFSI

  5. Isolation systems influence in the seismic loading propagation analysis applied to an innovative near term reactor

    International Nuclear Information System (INIS)

    Lo Frano, R.; Forasassi, G.

    2010-01-01

    Integrity of a Nuclear Power Plant (NPP) must be ensured during the plant life in any design condition and, particularly, in the event of a severe earthquake. To investigate the seismic resistance capability of as-built structures systems and components, in the event of a Safe Shutdown Earthquake (SSE), and analyse its related effects on a near term deployment reactor and its internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was applied considering, also, the use of innovative anti-seismic techniques. In this paper the attention is focused on the use and influence of seismic isolation technologies (e.g. isolators based on passive energy dissipation) that seem able to ensure the full integrity and operability of NPP structures, to enhance the seismic safety (improving the design of new NPPs and if possible, to retrofit existing facilities) and to attain a standardization plant design. To the purpose of this study a numerical assessment of dynamic response/behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) are herein presented and compared in order to highlight the isolation technique effectiveness.

  6. Seismic evaluation of existing nuclear power plants

    International Nuclear Information System (INIS)

    2003-01-01

    programmes at existing operating plants are plant specific or regulatory specific. This means that this report is meant to define the minimum generic requirements and may need to be supplemented on a plant specific basis to consider particular aspects of the original design basis. Among the options available, two methods are particularly appropriate for assessing the seismic safety of facilities, the seismic margin assessment (SMA) method and the seismic probabilistic safety assessment (SPSA) method. Both SMA and SPSA are discussed in this report. Current NPP design criteria and comprehensive seismic design procedures, as applied to the design of new facilities but using a re-evaluated seismic input, may be applied in the seismic evaluation programme. It is noted that these would be a conservative and usually expensive approach for evaluation of an existing operating facility and they are not discussed further in this report. Evaluation of existing NPPs may result in the identification of items of the SSSC list which have to be upgraded. Upgrading itself is not covered by this Safety Report; however, some general principles are presented in order to preserve consistency between evaluation and upgrading processes. (It should be pointed out that when an upgrading programme has to be carried out, it necessitates more engineering resources than the evaluation process does; similarly upgrading is too large and complex a matter to be covered by this Safety Report.) Section 2 presents the general philosophy of seismic evaluation; Section 3 discusses data collection and investigations; Section 4 is devoted to seismic hazard assessment; Section 5 discusses the safety analysis of the NPP; Section 6 discusses the practice of walkdown; Section 7 covers the criteria and methods used for seismic capacity assessment of SSSCs; Section 8 discusses the principle of the design of a possible seismic upgrading; Section 9 specifies some rules of quality assurance and organization

  7. Defence in Depth by Design for the Advanced GIII NPP in China

    Energy Technology Data Exchange (ETDEWEB)

    Liu, S.; Zhang, Y.; Zhang, X., E-mail: liusongtao.npic@gmail.com [Science and Technology on Reactor System Design Technology Laboratory Chengdu, Sichuan (China)

    2014-10-15

    This paper describes the design of the advanced nuclear power plant ACP1000 in China that keeps the principle of defence in depth. To enhance the safety of the new generation NPPs, passive and active engineering safety features are used. The reactor will be kept safe under design basis accidents by using active engineering safety features, such as the medium and low pressure safety injection systems, and the emergency feedwater system. Under beyond DBAs, the passive safety systems will be actuated to keep removing residual heat for more than 72 hours, and to keep the core melt retained and cooled in the vessel. After the Fukushima nuclear accident, there are six main design enhancements in ACP1000 to meet the demands of the China authorities. (author)

  8. Formulation of the task on ergonomic designing of NPP operator activity

    International Nuclear Information System (INIS)

    Anokhin, A.N.

    1996-01-01

    One of the main causes of inefficiency of existing nuclear plant operator activity support means is the absence of common integrated system approach to ergonomic designing of operator activity. Some attempt to formalize the problem as a task of macro-ergonomic designing is made. The structure of anthropocentric functional model of human-operator-nuclear plant system operation is described. Operator activity is characterized by some resulting properties (such as reliability, etc.). These properties are influenced by human-operator internal properties and working environment external properties. The detailed classification of all these properties is offered. The main result of this work is the statement of tasks of operator activity macro-ergonomic designing based on the offered formalization

  9. A performance goal-based seismic design philosophy for waste repository facilities

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1994-02-01

    A performance goal-based seismic design philosophy, compatible with DOE's present natural phenomena hazards mitigation and ''graded approach'' philosophy, has been proposed for high level nuclear waste repository facilities. The rationale, evolution, and the desirable features of this method have been described. Why and how the method should and can be applied to the design of a repository facility are also discussed

  10. Performance-based seismic design of steel frames utilizing colliding bodies algorithm.

    Science.gov (United States)

    Veladi, H

    2014-01-01

    A pushover analysis method based on semirigid connection concept is developed and the colliding bodies optimization algorithm is employed to find optimum seismic design of frame structures. Two numerical examples from the literature are studied. The results of the new algorithm are compared to the conventional design methods to show the power or weakness of the algorithm.

  11. The Canarian Seismic Monitoring Network: design, development and first result

    Science.gov (United States)

    D'Auria, Luca; Barrancos, José; Padilla, Germán D.; García-Hernández, Rubén; Pérez, Aaron; Pérez, Nemesio M.

    2017-04-01

    Tenerife is an active volcanic island which experienced several eruptions of moderate intensity in historical times, and few explosive eruptions in the Holocene. The increasing population density and the consistent number of tourists are constantly raising the volcanic risk. In June 2016 Instituto Volcanologico de Canarias started the deployment of a seismological volcano monitoring network consisting of 15 broadband seismic stations. The network began its full operativity in November 2016. The aim of the network are both volcano monitoring and scientific research. Currently data are continuously recorded and processed in real-time. Seismograms, hypocentral parameters, statistical informations about the seismicity and other data are published on a web page. We show the technical characteristics of the network and an estimate of its detection threshold and earthquake location performances. Furthermore we present other near-real time procedures on the data: analysis of the ambient noise for determining the shallow velocity model and temporal velocity variations, detection of earthquake multiplets through massive data mining of the seismograms and automatic relocation of events through double-difference location.

  12. Contrast of aseismic design for NPP pressure pipelines of class 2

    International Nuclear Information System (INIS)

    Bai Wenting; Dai Junwu; Feng Guozhong; Rong Feng

    2011-01-01

    At present, the RCC-M of French, ASME (2007) of U.S.A and GB50267-97 of China are the primary nuclear technical codes for the design of facilities, systems, and components, which are similar in the classification of nuclear facilities in nu clear power plants, but are not exactly the same in the piping design clauses of the class 2 pressure pipeline. For the earthquake input methods, GB, ASME and RCC-M are basically similar. The hard soil standard response spectrum of GB is relatively safer. RCC-M emphasizes on the impact of the pressure, the GB50267-97 and ASME emphasize more on the weight and other occasional loads such as earthquakes effects. RCC-M is safer than GB and ASME in level D criteria. Case analysis shows that in the conditions of lower pressure and the same stress intensification factor, GB and ASME criteria are safer than RCC-M. (authors)

  13. Application of ecological interface design in nuclear power plant (NPP operator support system

    Directory of Open Access Journals (Sweden)

    Alexey Anokhin

    2018-05-01

    Full Text Available Most publications confirm that an ecological interface is a very efficient tool to supporting operators in recognition of complex and unusual situations and in decision-making. The present article describes the experience of implementation of an ecological interface concept for visualization of material balance in a drum separator of RBMK-type NPPs. Functional analysis of the domain area was carried out and revealed main factors and contributors to the balance. The proposed ecological display was designed to facilitate execution of the most complicated cognitive operations, such as comparison, summarizing, prediction, etc. The experimental series carried out at NPPs demonstrated considerable reduction of operators' mental load, time of reaction, and error rate. Keywords: Ecological Interface Design, Experimental Evaluation, Model, Work Domain Analysis

  14. NPP post-accident monitoring system based on unmanned aircraft vehicle:concept, design principles

    International Nuclear Information System (INIS)

    Sachenko, A.A.; Kochan, V.V.; Kharchenko, V.S.; Yanovskij, M.Eh.; Yastrebenetskij, M.A.; Fesenko, G.V.

    2016-01-01

    The paper presents a concept of designing the post-accident system for monitoring the equipment and territory of nuclear power plant after a severe accident based on unmanned aircraft vehicle (UAVs). Wired power and communications networks are found out as the most vulnerable ones during the accident monitoring, and informativity, reliability and veracity are recognized as system basic parameters. It is proposed to equip measurement and control modules with backup wireless communication channels and deploy the repeaters network based on UAVs to ensure the informativity. Modules possess the backup power battery, and repeaters appear in the appropriate places after the accident to provide the survivability. Moreover, an optimization of UAVs' location is proposed according to the minimum energy consumption criterion. To ensure the veracity, it is expected to design the noise-immune protocol for message exchange and archiving and self-diagnostics of all system components

  15. Base Isolation for Seismic Retrofitting of a Multiple Building Structure: Design, Construction, and Assessment

    Directory of Open Access Journals (Sweden)

    Massimiliano Ferraioli

    2017-01-01

    Full Text Available The paper deals with the seismic retrofit of a multiple building structure belonging to the Hospital Centre of Avellino (Italy. At first, the paper presents the preliminary investigations, the in situ measurements and laboratory tests, and the seismic assessment of the existing fixed-base structures. Having studied different strategies, base isolation proved to be the more appropriate, also for the possibility offered by the geometry of the building to easily create an isolation interface at the ground level. The paper presents the design project, the construction process, and the details of the isolation intervention. Some specific issues of base isolation for seismic retrofitting of multiple building structures were lightened. Finally, the seismic assessment of the base-isolated building was carried out. The seismic response was evaluated through nonlinear time-history analysis, using the well-known Bouc-Wen model as the constitutive law of the isolation bearings. For reliable dynamic analyses, a suite of natural accelerograms compatible with acceleration spectra of Italian Code was first selected and then applied along both horizontal directions. The results were finally used to address some of the critical issues of the seismic response of the base-isolated multiple building structure: accidental torsional effects and potential poundings during strong earthquakes.

  16. A Conceptual Design of Light-weighted Mobile Robot for the Integrity of SG Tubes in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Seung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Chun Sup; Park, Ki Tae [Korea Plant Serviceand Engineering, Busan (Korea, Republic of)

    2010-10-15

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or tube sheets of SG. Confusions for the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and it leads to the increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light-weighed mobile robots have been introduced by Westinghouse and Zetec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam-locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidently, they can be fall down and light repair works can be performed. This paper provides the design results for a light weighted mobile robot which is recently being developed in cooperation of our institutes

  17. A Conceptual Design of Light-weighted Mobile Robot for the Integrity of SG Tubes in NPP

    International Nuclear Information System (INIS)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Seung Ho; Shin, Chun Sup; Park, Ki Tae

    2010-01-01

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or tube sheets of SG. Confusions for the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and it leads to the increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light-weighed mobile robots have been introduced by Westinghouse and Zetec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam-locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidently, they can be fall down and light repair works can be performed. This paper provides the design results for a light weighted mobile robot which is recently being developed in cooperation of our institutes

  18. Seismic design and evaluation criteria for DOE facilities (DOE-STD-1020-XX)

    International Nuclear Information System (INIS)

    Short, S.A.; Kennedy, R.P.; Murray, R.C.

    1993-01-01

    Seismic design and evaluation criteria for DOE facilities are provided in DOE-STD-1020-XX. The criteria include selection of design/evaluation seismic input from probabilistic seismic hazard curves combined with commonly practiced deterministic response evaluation methods and acceptance criteria with controlled levels of conservatism. Conservatism is intentionally introduced in specification of material strengths and capacities, in the allowance of limited inelastic behavior and by a seismic load factor. These criteria are based on the performance or risk goals specified in DOE 5480.28. Criteria have been developed following a graded approach for several performance goals ranging from that appropriate for normal-use facilities to that appropriate for facilities involving hazardous or critical operations. Performance goals are comprised of desired behavior and of the probability of not achieving that behavior. Following the seismic design/evaluation criteria of DOE-STD-1020-XX is sufficient to demonstrate that the probabilistic performance or risk goals are achieved. The criteria are simple procedures but with a sound, rigorous basis for the achievement of goals

  19. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  20. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  1. Awarable complexity: a study on CRT picture design based on plant images by NPP operators

    International Nuclear Information System (INIS)

    Kawano, Ryutaro; Ohtsuka, Tsutomu; Masugi, Tsuyoshi

    2000-01-01

    Original pictures installed in the 1st and 2nd generation type central control panels (CCP) and new 'Awarable and Complex' pictures were made on personal computers and evaluated. A total 18 of actual plant operators (M=32.3, SD=10.5 years old) participated in the evaluation. The operators rated the new CRT pictures highly. The response times using the new CRT pictures were shorter than those by the original pictures. Both results suggested that the CRT picture design guidelines based on the operators' plant images were effective for improving their performance. (author)

  2. Research on the NPP human factors engineering operating experience review

    International Nuclear Information System (INIS)

    Ren Xiangchen; Miao Hongxing; Ning Zhonghe

    2006-01-01

    This paper addresses the importance of the human factors engineering (HFE) for the design of nuclear power plant (NPP), especially for the design of human-machine interface in the NPP. It also summarizes the scope and content of the NPP HFE. The function, scope, content and process of the NPP human factors engineering operating experience review (OER) are mainly focused on, and significantly discussed. Finally, it briefly introduces the situation of the studies on the OER in China. (authors)

  3. Designing in seismic areas in the third millennium: modern technologies

    International Nuclear Information System (INIS)

    Martelli, Alessandro

    2015-01-01

    The World Conference on Seismic Isolation, Energy Dissipation and Active Vibrations Control of Structures, which took place in Sendai (Japan) on September 24-26, 2013. Other papers presented at this conference deal with the use of the traditional approach. More updated information on the application of the AS systems became available at the ASSISi 14. World Conference, held in San Diego (California, USA) on September 7-11, 2015. Most SI systems rely on the use of rubber bearings (RBs), such as the High Damping natural Rubber Bearings (HDRBs), Neoprene Bearings (NBs), Lead Rubber Bearings (LRBs), or (especially in Japan) Low Damping Rubber Bearings (LDRBs) in parallel with dampers; in buildings, some plane surfaces steel-Teflon (PTFE) Sliding Devices (SDs) are frequently added to the RBs to support their light parts without unnecessarily stiffening the SI system (which would make it less effective) and (if they are significantly asymmetric in the horizontal plane) to minimize the torsion effects (the effects of the vertical asymmetries are drastically reduced by the quasi 'rigid body motion' of the seismically isolated superstructure). Another type of isolators, which has been used in Italy after the 2009 Abruzzo earthquake, is the so-called Curved Surface Slider (CSS), which derived from the US Friction Pendulum (FPS) and the subsequent German Seismic Isolation Pendulum (SIP). Finally, rolling isolators (in particular Ball Bearings, BBs, and Sphere Bearings) are also applied: they are very effective and find numerous applications (more than 200 in 2013) to protect buildings in Japan, but not in Italy, because there they have been judged to be too expensive (however, they have already been used, even in Italy, to protect precious masterpieces and other contents of museums, as well as costly equipment, including that of operating-rooms in hospitals). It shall be stressed that, to the knowledge of the author, all structures protected by RBs that were located

  4. Operation Aspect of the Main Control Room of NPP

    International Nuclear Information System (INIS)

    Sahala M Lumbanraja

    2009-01-01

    The main control room of Nuclear Power Plant (NPP) is operational centre to control all of the operation activity of NPP. NPP must be operated carefully and safely. Many aspect that contributed to operation of NPP, such as man power whose operated, technology type used, ergonomic of main control room, operational management, etc. The disturbances of communication in control room must be anticipated so the high availability of NPP can be achieved. The ergonomic of the NPP control room that will be used in Indonesia must be designed suitable to anthropometric of Indonesia society. (author)

  5. Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

    International Nuclear Information System (INIS)

    Blom, F.J.

    2007-01-01

    Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

  6. Safety Analysis in Design and Assessment of the Physical Protection of the OKG NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lindahl, P., E-mail: par.lindahl@okg.eon.se [OKG Aktiebolag, Oskarshamn (Sweden)

    2014-10-15

    OKG AB operates a three unit nuclear power plant in the southern parts of Sweden. As a result of recent development of the legislation regarding physical protection of nuclear facilities, OKG has upgraded the protection against antagonistic actions. The new legislation includes requirements both on specific protective measures and on the performance of the physical protection as a whole. In short, the performance related requirements state that sufficient measures shall be implemented to protect against antagonistic actions, as defined by the regulator in the “Design Basis Threat” (DBT). Historically, physical protection and nuclear safety has been managed much as separate issues with different, sometimes contradicting, objectives. Now, insights from the work with the security upgrade have emphasized that physical protection needs to be regarded as an important part of the Defence-In-Depth (DiD) against nuclear accidents. Specifically, OKG has developed new DBT-based analysis methods, which may be characterized as probabilistically informed deterministic analysis, conformed to a format similar to the one used for conventional internal events analysis. The result is a powerful tool for design and assessment of the performance of the protection against antagonistic actions, using a nuclear safety perspective. (author)

  7. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  8. Estimation of Cyclic Interstory Drift Capacity of Steel Framed Structures and Future Applications for Seismic Design

    Directory of Open Access Journals (Sweden)

    Edén Bojórquez

    2014-01-01

    Full Text Available Several studies have been devoted to calibrate damage indices for steel and reinforced concrete members with the purpose of overcoming some of the shortcomings of the parameters currently used during seismic design. Nevertheless, there is a challenge to study and calibrate the use of such indices for the practical structural evaluation of complex structures. In this paper, an energy-based damage model for multidegree-of-freedom (MDOF steel framed structures that accounts explicitly for the effects of cumulative plastic deformation demands is used to estimate the cyclic drift capacity of steel structures. To achieve this, seismic hazard curves are used to discuss the limitations of the maximum interstory drift demand as a performance parameter to achieve adequate damage control. Then the concept of cyclic drift capacity, which incorporates information of the influence of cumulative plastic deformation demands, is introduced as an alternative for future applications of seismic design of structures subjected to long duration ground motions.

  9. Analytical Study on the Beyond Design Seismic Capacity of Reinforced Concrete Shear Walls

    Energy Technology Data Exchange (ETDEWEB)

    Nugroho, Tino Sawaldi Adi [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chi, Ho-Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The OECD-NEA has organized an international benchmarking program to better understand this critical issue. The benchmark program provides test specimen geometry, test setup, material properties, loading conditions, recorded measures, and observations of the test specimens. The main objective of this research is to assess the beyond design seismic capacity of the reinforced concrete shear walls tested at the European Laboratory for Structural Assessment between 1997 and 1998 through participation in the OECD-NEA benchmark program. In this study, assessing the beyond design seismic capacity of reinforced concrete shear walls is performed analytically by comparing numerical results with experimental results. The seismic shear capacity of the reinforced concrete shear wall was predicted reasonably well using ABAQUS program. However, the proper calibration of the concrete material model was necessary for better prediction of the behavior of the reinforced concrete shear walls since the response was influenced significantly by the material constitutive model.

  10. Seismic design principles for the German fast breeder reactor SNR 2

    International Nuclear Information System (INIS)

    Busch, K.A.; Peters, K.A.; Rosenhauer, W.

    1987-01-01

    The safety issue of an adequate and optimized external event protection is of course that unnecessary hardware precautions might promote internal disturbances or hamper their control. It has up to now not satisfactorily been realized that the only serious context for seismic impacts on a fast reactor is their attributed potential of overriding core disruptive accident prevention, see e.g. GRS 1982. General and exaggerated antiseismic design features not focussed upon this point may as well turn out to be non-negligible initators in the absence of seismic vibrations. Unexpected snubber difficulties requiring additional reactor scrams and decay heat removal phases may be named as a simple example. The presented seismic design principles reflect the progress made in the concerned fields of analysis and do serve on the other hand as guidelines for research and development efforts under work. (orig./GL)

  11. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

  12. Hydrogen-management in beyond design accident conditions in NPP Neckar 2

    International Nuclear Information System (INIS)

    Zaiss, W.

    1999-01-01

    Neckar 2 is a 1340 MWE 4-loop pressurized water reactor (PWR) of Siemens KONVOI type, located in the south of Germany. It was first connected to the grid in January 1989. Commercial operation started in April 1989. Task assignment: In Germany it was recommended by the Reactor Safety Commission (RSK) on December 17, 1997, to reequip passive autocatalytic recombiners for the controlling of the hydrogen problem. The removal of the hydrogen is an essential part which guarantees the integrity of the containment. The implementation of the recombiners is a further step for the decrease of the nuclear rest risk. The RSK confirmed, that the implementation of the passive autocatalytic recombiners is a safety measure for the controlled removal of the hydrogen in beyond design accident conditions. Assumption : Failure of the whole residual heat removal system (RHRS) and non sufficient effect of the systems which have been installed for beyond design accident conditions. Effect on the reactor coolant system (RCS): The reactor core will be damaged by non sufficient cooling with the output of hydrogen because all the specified emergency actions have failed. The overheating of the core is responsible for the production of hydrogen by the reaction of zirconium of the fuel-rod cladding with the water vapour. In case of nuclear superheating it would be possible that the reactor vessel would start smelting. The interacting between the core and the concrete, together with the armouring of the biological shield would also produce hydrogen. The hydrogen would escape together with the water vapour out of the leak and would spread out into the whole containment. Results : the number and the position of the different sized recombiners were determined on engineering judgement. the following 4 scenarios are representatively. The 4 scenarios were analyzed for in beyond design accident conditions with the MELCOR-Code: No. 1: Loss of main feedwater supply with primary feed and bleed. No. 2

  13. The 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1995-01-01

    A record of the 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping is provided. The focus of the forum was the design-by-rule method for seismic design of piping. Issues such as acceptance criteria, ductility considerations, demonstration of margin, training, verification and costs were discussed. The use of earthquake experience data, including the recent Northridge earthquake, to justify a design-by-rule method was explored. The majority of the participants felt there are not significant advantages to developing a design-by-rule approach for new plant design. One major disadvantage was considered by many to be training. Extensive training will be required to properly implement a design-by-rule approach. Verification of designs was considered by the majority to be equally important for design-by-rule as for design-by-analysis. If a design-by-rule method is going to be effective, the method will have to be based on ductility considerations (UBC approach). A significant issue will be justification of seismic margins with liberal rules. The UBC approach is being questioned by some because of the recent structural cracking problems in the Northridge earthquake

  14. Pre elementary design of primary reformer for hydrogen plant coupled with HTGR type NPP

    International Nuclear Information System (INIS)

    Dedy Priambodo; Erlan Dewita; Sudi Ariyanto

    2012-01-01

    Hydrogen has a high potent for new energy, because of it availability. Steam reforming is a fully developed commercial technology and is the most economical method for production of hydrogen. Steam reforming uses an external source of hot gas to heat tubes in which a catalytic reaction takes place that converts steam and lighter hydrocarbons such as natural gas (methane) or refinery feedstock into hydrogen and carbon monoxide (syngas) at high temperature on primary reformer (800-900°C). Utilization of helium from HTGR as heating medium for primary reformer has consequence to type and shape of its reactor. The main goal of this paper is to determine type/shape and pre elementary design of chemical reactor for the cogeneration system of Hydrogen Plant and HTGR The primary reformer for this system is Fixed Bed Multitube reactor with specification tube: NPS 3,5 Sch 40 ST 40S, 0.281 in thickness, number of tube 849 pieces and ASTM HH 30 for tube material. Tube arrangement is 'triangular pitch' on shell Split-Ring Floating Head from Steel Alloy SA 301 Grade B equipted with 8 baffles. (author)

  15. Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-01-01

    The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, components for which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation

  16. Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Stevenson, J D [Stevenson and Associates, Cleveland, OH (United States)

    1995-07-01

    The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, componentsfor which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation.

  17. The Role of Tectonic and Seismicity in Siting of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Abdel Aziz, M.A.H.

    2008-01-01

    The site selection for the a nuclear power plant (NPP) is controlled by many criteria. One of the most important criterion is the tectonic and seismicity of the site and its surroundings. Since, it is preferable the site in concern is characterized by low tectonic and low seismicity to avoid the damage effects associated with the occurrence of destructive earthquakes. The investigation of the tectonic and seismicity maps of egypt has been carried out to candidate potential areas or sites for nuclear power plant installation from seismicity point of view. Also, the design basis ground motion in terms of peak ground acceleration and response spectra of some of the potential sites are defined through the conduct of probabilistic seismic hazard analysis The study revealed that although there is no criterion to exclude areas of high tectonic and high seismicity as potential sites for nuclear power plant installation but, it is preferable avoiding such areas. This is attributed to the critical seismic curve that characterizes such areas and is required high seismic design levels to resist the destructive vibratory ground motion associated with the expected earthquake. Consequently, the required high seismic design levels will have a negative impact on the economic cost of the facility compared with that built in low and moderate seismic areas. Hence, areas like the gulf of suez, the northern part of the Red Sea and the southern part of Sinai Peninsula should be avoided as potential sites for NPP from the tectonic and seismicity point of view. On the other hand, areas like Nile delta and its valley, the Northern and Southern parts of Western desert and the central and southern parts of the Eastern Desert should be candidate as potential sites on condition, the other criteria meet the IAEA's regulations. Also, the seismic hazard curve of the Northwest littoral zone reflects low design basis ground motion values compared with the Nile delta region

  18. Multi Canister Overpack (MCO) Handling Machine Trolley Seismic Uplift Constraint Design Loads

    International Nuclear Information System (INIS)

    SWENSON, C.E.

    2000-01-01

    The MCO Handling Machine (MHM) trolley moves along the top of the MHM bridge girders on east-west oriented rails. To prevent trolley wheel uplift during a seismic event, passive uplift constraints are provided as shown in Figure 1-1. North-south trolley wheel movement is prevented by flanges on the trolley wheels. When the MHM is positioned over a Multi-Canister Overpack (MCO) storage tube, east-west seismic restraints are activated to prevent trolley movement during MCO handling. The active seismic constraints consist of a plunger, which is inserted into slots positioned along the tracks as shown in Figure 1-1. When the MHM trolley is moving between storage tube positions, the active seismic restraints are not engaged. The MHM has been designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis (Reference 3) reported seismic uplift restraint loading and EDERER performed corresponding structural calculations. The ALSTHOM and EDERER calculations were performed with the east-west seismic restraints activated and the uplift restraints experiencing only vertical loading. In support of development of the CSB Safety Analysis Report (SAR), an evaluation of the MHM seismic response was requested for the case where the east-west trolley restraints are not engaged. For this case, the associated trolley movements would result in east-west lateral loads on the uplift constraints due to friction, as shown in Figure 1-2. During preliminary evaluations, questions were raised as to whether the EDERER calculations considered the latest ALSTHOM seismic analysis loads (See NCR No. 00-SNFP-0008, Reference 5). Further evaluation led to the conclusion that the EDERER calculations used appropriate vertical loading, but the uplift restraints would need to be re-analyzed and modified to account for lateral loading. The disposition of NCR 00-SNFP-0008 will track the redesign and modification effort. The purpose of this calculation is to establish bounding seismic

  19. Preclosure seismic design methodology for a geologic repository at Yucca Mountain. Revision 1

    International Nuclear Information System (INIS)

    1996-08-01

    This topical report is the second in a series of three reports being developed by the US Department of Energy (DOE) to document the preclosure seismic design of structures, systems, and components (SSCs) that are important to the radiological safety of the potential repository at Yucca Mountain, Nevada. The first topical report, Methodology to Assess Fault Displacement and Vibratory Ground Motion Hazards at Yucca Mountain, YMP/TR-002-NP, was submitted to the US Nuclear Regulatory Commission (NRC) staff for review and comment in 1994 and has been accepted by the staff. The DOE plans to implement this methodology in fiscal year 1997 to develop probabilistic descriptions of the vibratory ground motion hazard and the fault displacement hazard at the Yucca Mountain site. The second topical report (this report) describes the DOE methodology and acceptance criteria for the preclosure seismic design of SSCs important to safety. A third report, scheduled for fiscal year 1998, will document the results of the probabilistic seismic hazard assessment (conducted using the methodology in the first topical report) and the development of the preclosure seismic design inputs. This third report will be submitted to NRC staff for review and comment as a third topical report or as a design study report

  20. Numerical Investigation of Progressive Collapse Resistance for Seismically Designed RC Buildings

    OpenAIRE

    Marchiş, Adrian G.; Ioani, Adrian M.

    2014-01-01

    In this paper the progressive collapse behavior of a reinforced concrete framed building located in different seismic areas from Romania is investigated. The six-storey structure is designed for low (ag = 0.08 g), moderate (ag = 0.16 g) and high (ag = 0.24 g) seismic zone. Based on the GSA (2003) criteria, a nonlinear static analysis is conducted first in order to estimate the progressive collapse resistance of the models. It was shown that all the structures will collapse when subjected to i...

  1. Seismic demand evaluation based on actual earthquake records

    International Nuclear Information System (INIS)

    Jhaveri, D.P.; Czarnecki, R.M.; Kassawara, R.P.; Singh, A.

    1990-01-01

    Seismic input in the form of floor response spectra (FRS) are needed in seismic design and evaluation of equipment in nuclear power plants (NPPs). These are typically determined by analytical procedures using mathematical models of NPP structures and are known to be very conservative. Recorded earthquake data, in the form of acceleration response spectra computed from the recorded acceleration time histories, have been collected from NPP structures located in seismically active areas. Statistics of the ratios, or amplification factors, between the FRS at typical floors and the acceleration response spectra at the basemat or in the freefield, are obtained for typical NPP structures. These amplification factors are typically in terms of the peak spectral and zero period values, as well as a function of frequency. The average + 1σ values of these ratios, for those cases where enough data are available, are proposed to be used as limits to analytically calculated FRS, or for construction of simplified FRS for determining seismic input or demand in equipment qualification. (orig.)

  2. Consideration on the applicability of the design seismic coefficient of a large cutting slope under the strong earthquake

    International Nuclear Information System (INIS)

    Ito, Hiroshi; Sawada, Yoshihiro; Satou, Kiyotaka

    1989-01-01

    In this study, the characteristic of equivalent seismic coefficient and the applicability of the design seismic coefficient of a large cutting rock slope around Nuclear Power Plant were examined by analytical parameter survey. As the results, the equivalent seismic coefficient by dynamic analysis become great with increase of transverse elastic wave velocity and the case of long period motion. That is, as the wave length of rock mass become longer, the equivalent seismic coefficient become great parabolically. Moreover, there is a inverse proportion relation between the ratio (dynamic safety factor/static safety factor) and wave length. In addition, the graph to forecast the dynamic sliding safety factor under the input seismic motion of the max. Acceleration 500 gal from the result of static simple method was proposed and the applicable range of design seismic coefficient of rock slope was indicated. (author)

  3. Organization of monitoring of agricultural products in NPP region

    International Nuclear Information System (INIS)

    Panteleev, L.I.; Spirin, E.V.; Sanzharova, N.I.

    1990-01-01

    Problem of organizing chemical and radiation monitoring of agricultural products in NPP region is considered. Attention is paid to monitoring during NPP siting and designing, to monitoring of radioactive contamination of agricultural products under normal NPP operation, emergency situations and decommissioning

  4. Decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    Anon

    2009-01-01

    In this presentation the Operation history of A1 NPP, Project 'Decommissioning of A1 NPP' - I stage, Project 'Decommissioning of A1 NPP ' - II stage and Next stages of Project 'Decommissioning of A1 NPP ' are discussed.

  5. Development of seismic design method for free standing rack and applicability to Japanese nuclear power plant

    International Nuclear Information System (INIS)

    Takaki, Yu; Taniguchi, Katsuhiko; Kishimoto, Junichi; Iwasaki, Akihisa; Nekomoto, Yoshitsugu; Kuga, Tohru; Kameyama, Masashi

    2017-01-01

    Free standing racks which are not anchored to the pool floor nor walls have never been adopted in Japan. Under an earthquake, behaviors of free standing racks are nonlinear and involve a complex combination of motions (sliding, rocking, and twisting) and impacts between a fuel assembly and the fuel cell walls and between a pit floor and rack pedestals. To predict a seismic response of free standing racks, the seismic analysis requires careful considerations of these complex phenomena (sliding, rocking, and twisting), fluid coupling effects and frictional effects. We carried out seismic experiments on the full-scale rack model in both water and dry conditions and obtained the fundamental data about behavior of free standing racks (sliding, and rocking motions). We have developed the nonlinear dynamic analysis method to predict seismic response of free standing racks utilizing the full-scale test result and verified the analysis evaluation method of free standing rack by comparison between analysis results and experimental data. Furthermore, we applied the seismic design method to the free standing rack in the Japanese nuclear plant (Mihama nuclear power station Unit 3), and verified that the free standing rack was applicable to Japanese nuclear plant. (author)

  6. The Design of Wireless Data Acquisition and Remote Transmission Interface in Micro-seismic Signals

    Directory of Open Access Journals (Sweden)

    Huan-Huan BIAN

    2014-02-01

    Full Text Available The micro-seismic signal acquisition and transmission is an important key part in geological prospecting. This paper describes a bran-new solution of micro-seismic signal acquisition and remote transmission using Zigbee technique and wireless data transmission technique. The hardware such as front-end data acquisition interface made up by Zigbee wireless networking technique, remote data transmission solution composed of general packet radio service (or GPRS for short technique and interface between Zigbee and GPRS is designed in detail. Meanwhile the corresponding software of the system is given out. The solution solves the numerous practical problems nagged by complex and terrible environment faced using micro-seismic prospecting. The experimental results demonstrate that the method using Zigbee wireless network communication technique GPRS wireless packet switching technique is efficient, reliable and flexible.

  7. Subsurface Characterization using Geophysical Seismic Refraction Survey for Slope Stabilization Design with Soil Nailing

    Science.gov (United States)

    Ashraf Mohamad Ismail, Mohd; Ng, Soon Min; Hazreek Zainal Abidin, Mohd; Madun, Aziman

    2018-04-01

    The application of geophysical seismic refraction for slope stabilization design using soil nailing method was demonstrated in this study. The potential weak layer of the study area is first identify prior to determining the appropriate length and location of the soil nail. A total of 7 seismic refraction survey lines were conducted at the study area with standard procedures. The refraction data were then analyzed by using the Pickwin and Plotrefa computer software package to obtain the seismic velocity profiles distribution. These results were correlated with the complementary borehole data to interpret the subsurface profile of the study area. It has been identified that layer 1 to 3 is the potential weak zone susceptible to slope failure. Hence, soil nails should be installed to transfer the tensile load from the less stable layer 3 to the more stable layer 4. The soil-nail interaction will provide a reinforcing action to the soil mass thereby increasing the stability of the slope.

  8. Recent results of a seismically isolated optical table prototype designed for advanced LIGO

    International Nuclear Information System (INIS)

    Sannibale, V; Abbott, B; Boschi, V; Coyne, D; DeSalvo, R; Aso, Y; Marka, S; Ottaway, D; Stochino, A

    2008-01-01

    The Horizontal Access Module Seismic Attenuation System (HAM-SAS) is a mechanical device expressly designed to isolate a multipurpose optical table and fit in the tight space of the LIGO HAM Ultra-High-Vacuum chamber. Seismic attenuation in the detectors' sensitivity frequency band is achieved with state of the art passive mechanical attenuators. These devices should provide an attenuation factor of about 70dB above 10Hz at the suspension point of the Advanced LIGO triple pendulum suspension. Automatic control techniques are used to position the optical table and damp rigid body modes. Here, we report the main results obtained from the full scale prototype installed at the MIT LIGO Advanced System Test Interferometer (LASTI) facility. Seismic attenuation performance, control strategies, improvements and limitations are also discussed

  9. Seismic soil–structure interaction analysis of a nuclear power plant building founded on soil and in degraded concrete stiffness condition

    International Nuclear Information System (INIS)

    Farahani, Reza V.; Dessalegn, Tewodros M.; Vaidya, Nishikant R.; Bazan-Zurita, Enrique

    2016-01-01

    . Therefore, it is important to consider SSI parameter variability in the seismic design or evaluation of NPP buildings that bear on soil.

  10. Making the Most of the Operating Experience Feedback: Design of a Computerized Data Base Related to NPP Component Behavior

    International Nuclear Information System (INIS)

    Degrave, C.; Martin-Onraet, M.

    1998-01-01

    Electricite de France carries out several analyses on NPP component behavior every year. However, these analyses are conducted separately by specialists of only the kind of component and regardless of the time factor. Seven years ago, a working group from the Nuclear Operation Division completed a detailed examination of component behavior of the 50 NPP then in operation, over a ten-years period (from 1980 to 1990). These analyses have been centered on main components only. Today, another working group will integrate this operating experience in a computerized process to facilitate access to all the information contained in the data base. This system will be based on a site server allowing a user-friendly connection, and updated periodically. (author)

  11. Design and Implementation of the National Seismic Monitoring Network in the Kingdom of Bhutan

    Science.gov (United States)

    Ohmi, S.; Inoue, H.; Chophel, J.; Pelgay, P.; Drukpa, D.

    2017-12-01

    Bhutan-Himalayan district is located along the plate collision zone between Indian and Eurasian plates, which is one of the most seismically active region in the world. Recent earthquakes such as M7.8 Gorkha Nepal earthquake in April 25, 2015 and M6.7 Imphal, India earthquake in January 3, 2016 are examples of felt earthquakes in Bhutan. However, there is no permanent seismic monitoring system ever established in Bhutan, whose territory is in the center of the Bhutan-Himalayan region. We started establishing permanent seismic monitoring network of minimum requirements and intensity meter network over the nation. The former is composed of six (6) observation stations in Bhutan with short period weak motion and strong motion seismometers as well as three (3) broad-band seismometers, and the latter is composed of twenty intensity meters located in every provincial government office. Obtained data are transmitted to the central processing system in the DGM office in Thimphu in real time. In this project, DGM will construct seismic vault with their own budget which is approved as the World Bank project, and Japan team assists the DGM for site survey of observation site, designing the observation vault, and designing the data telemetry system as well as providing instruments for the observation such as seismometers and digitizers. We already started the operation of the six (6) weak motion stations as well as twenty (20) intensity meter stations. Additionally, the RIMES (Regional Integrated Multi-hazard Early Warning System for Africa and Asia) is also providing eight (8) weak motion stations and we are keeping close communication to operate them as one single seismic monitoring network composed of fourteen (14) stations. This network will be definitely utilized for not only for seismic disaster mitigation of the country but also for studying the seismotectonics in the Bhutan-Himalayan region which is not yet precisely revealed due to the lack of observation data in the

  12. Earthquake response spectra for seismic design of nuclear power plants in the UK

    International Nuclear Information System (INIS)

    Bommer, Julian J.; Papaspiliou, Myrto; Price, Warren

    2011-01-01

    Highlights: → Seismic design of UK nuclear power plants usually based on PML response spectra. → We review derivation of PML spectra in terms of earthquake data used and procedure. → The data include errors and represent a small fraction of what is now available. → Seismic design loads in current practice are derived as mean uniform hazard spectra. → The need to capture epistemic uncertainty makes use of single equation indefensible. - Abstract: Earthquake actions for the seismic design of nuclear power plants in the United Kingdom are generally based on spectral shapes anchored to peak ground acceleration (PGA) values obtained from a single predictive equation. Both the spectra and the PGA prediction equation were derived in the 1980s. The technical bases for these formulations of seismic loading are now very dated if compared with the state-of-the-art in this field. Alternative spectral shapes are explored and the options, and the associated benefits and challenges, for generating uniform hazard response spectra instead of fixed shapes anchored to PGA are discussed.

  13. Philosophy for seismic design of nuclear power plants

    International Nuclear Information System (INIS)

    Teramae, Tetsuo

    1981-01-01

    In Japan, earthquakes occur frequently, therefore the basic philosophy in the aseismatic design of nuclear facilities is to design so as not to cause the accident which gives to the public in the surroundings and the employes radiation injuries in the case of large earthquakes. The ''Guideline for the aseismatic design techniques for nuclear power stations'' was drawn up in 1970 as the result of studies by related government offices and organizations. The guideline for determining the earthquakes used for design was published later, and the allowable stress for equipments and pipings has been adopted in accordance with ASME Code, Section 3. The buildings and structures, equipments and pipings in nuclear facilities are classified into three classes according to their importance in aseismatic design. The power of design earthquakes is determined corresponding to the degree of importance. The determination of the standard earthquake waves is explained. The proprieth of aseismatic design is evaluated on the basis of the basic concept of the combination of loads and the allowable limit. The static analysis in accordance with the Building Standards Act is applied to the B and C classes, while the dynamic analysis is required for the A class. The aseismatic analysis of buildings and structures, equipments and pipings is outlined. Many problems to be solved still remain though the concept of aseismatic design has been clarified. (Kako, I.)

  14. A performance goal-based seismic design philosophy for waste repository facilities

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1994-01-01

    A performance goal-based seismic design philosophy, compatible with DOE's present natural phenomena hazards mitigation and open-quotes graded approachclose quotes philosophy, has been proposed for high level nuclear waste repository facilities. The rationale, evolution, and the desirable features of this method have been described. Why and how the method should and can be applied to the design of a repository facility are also discussed

  15. Cernavoda NPP Knowledge Transfer

    International Nuclear Information System (INIS)

    Valache, C. M.

    2016-01-01

    Full text: The paper presents a description of the Knowledge Transfer (KT) process implemented at Cernavoda NPP, its designing and implementation. It is underlined that applying a KT approach should improve the value of existing processes of the organization through: • Identifying business, operational and safety risks due to knowledge gaps, • Transfer of knowledge from the ageing workforce to the peers and/or the organization, • Continually learning from successes and failures of individual or teams, • Convert tacit knowledge to explicit knowledge, • Improving operational and safety performance through creating both new knowledge and better access to existing knowledge. (author

  16. Seismic design of circular-section concrete-lined underground openings: Preclosure performance considerations for the Yucca Mountain Site

    International Nuclear Information System (INIS)

    Richardson, A.M.; Blejwas, T.E.

    1992-01-01

    Yucca Mountain, the potential site of a repository for high-level radioactive waste, is situated in a region of natural and man-made seismicity. Underground openings excavated at this site must be designed for worker safety in the seismic environment anticipated for the preclosure period. This includes accesses developed for site characterization regardless of the ultimate outcome of the repository siting process. Experience with both civil and mining structures has shown that underground openings are much more resistant to seismic effects than surface structures, and that even severe dynamic strains can usually be accommodated with proper design. This paper discusses the design and performance of lined openings in the seismic environment of the potential site. The types and ranges of possible ground motions (seismic loads) are briefly discussed. Relevant historical records of underground opening performance during seismic loading are reviewed. Simple analytical methods of predicting liner performance under combined in situ, thermal, and seismic loading are presented, and results of calculations are discussed in the context of realistic performance requirements for concrete-lined openings for the preclosure period. Design features that will enhance liner stability and mitigate the impact of the potential seismic load are reviewed. The paper is limited to preclosure performance concerns involving worker safety because present decommissioning plans specify maintaining the option for liner removal at seal locations, thus decoupling liner design from repository postclosure performance issues

  17. Mochovce NPP simulator

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    Mochovce NPP simulator basic features and detailed description of its characteristics are presented with its performance, certification and application for training of NPP operators as well as the training scenario

  18. Design of fuelling machine bridge and carriage to meet seismic qualification requirements

    International Nuclear Information System (INIS)

    Ghare, A.B.; Chhatre, A.G.; Vyas, A.K.; Bhambra, H.S.

    1996-01-01

    During each refuelling operation, the boundary of Primary heat transport system is extended up to Fuelling Machines. A breach in the pressure boundary of Fuelling Machine in this condition would cause a loss of coolant accident. Fuelling Machines are also used for transit storage of spent fuel bundles till discharged to fuel transfer system. Therefore, a fuelling machine, including its support structures, is required to be seismically qualified for both on-reactor ( coupled ) mode and off-reactor (uncoupled) mode. The fuelling machine carriage used in the first generation of Indian PHWRs is a mobile equipment on wheels moving over fixed rails. As this configuration was found unsuitable for withstanding strong seismic disturbances, a bridge type design with fixed columns was evolved for the next generation of reactors. Initially, the seismic analysis of the fuelling machine bridge and carriage was done using static structural analysis and values of natural frequencies for various structures were computed. The structures were suitably modified based on the results of this analysis. Subsequently, a detailed dynamic seismic analysis using finite element model has been completed for both coupled and uncoupled conditions. The qualification of the structure has been carried out as per ASME section 111 Division 1, sub section NF. Details of the significant design features, static and dynamic analysis, results and conclusions are given in the presentation. (author). 4 refs., 4 tabs., 7 figs

  19. Design of fuelling machine bridge and carriage to meet seismic qualification requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ghare, A B; Chhatre, A G; Vyas, A K; Bhambra, H S [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    During each refuelling operation, the boundary of Primary heat transport system is extended up to Fuelling Machines. A breach in the pressure boundary of Fuelling Machine in this condition would cause a loss of coolant accident. Fuelling Machines are also used for transit storage of spent fuel bundles till discharged to fuel transfer system. Therefore, a fuelling machine, including its support structures, is required to be seismically qualified for both on-reactor ( coupled ) mode and off-reactor (uncoupled) mode. The fuelling machine carriage used in the first generation of Indian PHWRs is a mobile equipment on wheels moving over fixed rails. As this configuration was found unsuitable for withstanding strong seismic disturbances, a bridge type design with fixed columns was evolved for the next generation of reactors. Initially, the seismic analysis of the fuelling machine bridge and carriage was done using static structural analysis and values of natural frequencies for various structures were computed. The structures were suitably modified based on the results of this analysis. Subsequently, a detailed dynamic seismic analysis using finite element model has been completed for both coupled and uncoupled conditions. The qualification of the structure has been carried out as per ASME section 111 Division 1, sub section NF. Details of the significant design features, static and dynamic analysis, results and conclusions are given in the presentation. (author). 4 refs., 4 tabs., 7 figs.

  20. Former Soviet Regulations for seismic design of NPPs and comparison with current international practice

    International Nuclear Information System (INIS)

    Kostarev, V.; Schukin, A.; Berkovski, A.

    1997-01-01

    This paper presents a summary of current earthquake design criteria used in former Soviet Regulations for equipment and piping systems of nuclear power plants in light of those used in United States and Japan. The detailed comparative seismic analysis of PWR (WWER) Primary Coolant Loop System (PCLS) according to Former Soviet (Russian) PNAE Code and ASME BPV Code with some comments regarding to Japan Code JEAG - 4601 was undertaken for better understanding of the differences and coincidences of seismic design criteria and requirements. The selection of these three guides for the study has very simple explanation: according to ASME BVPC, JEAG and PNAE the huge majority of existing NPPs has been designed. (J.P.N.)

  1. Former Soviet Regulations for seismic design of NPPs and comparison with current international practice

    Energy Technology Data Exchange (ETDEWEB)

    Kostarev, V; Schukin, A; Berkovski, A [CKTI-Vibroseism Co. Ltd. (Cape Verde)

    1997-03-01

    This paper presents a summary of current earthquake design criteria used in former Soviet Regulations for equipment and piping systems of nuclear power plants in light of those used in United States and Japan. The detailed comparative seismic analysis of PWR (WWER) Primary Coolant Loop System (PCLS) according to Former Soviet (Russian) PNAE Code and ASME BPV Code with some comments regarding to Japan Code JEAG - 4601 was undertaken for better understanding of the differences and coincidences of seismic design criteria and requirements. The selection of these three guides for the study has very simple explanation: according to ASME BVPC, JEAG and PNAE the huge majority of existing NPPs has been designed. (J.P.N.)

  2. Towards safe and economic seismic design of cooling towers of extreme height

    International Nuclear Information System (INIS)

    Kraetzig, W.B.; Meskouris, K.

    1979-01-01

    Nuclear power plants are being increasingly equipped with natural draught cooling towers of heights greater than 160 m. In many arid zones, where high natural draught cooling towers with dry cooling systems are being projected, wind loads are relativelly small while site seismicity is relatively high. Thus the ability of the tower to withstand earthquake induced forces governs its design. On the other hand, most reinforced concrete cooling towers of extreme height built so far were designed to withstand high wind loads and moderate earthquake loads. The effects of special structural measures for obtaining an economic design, such as the introduction of ring stiffened shells, have been studied mainly for those towers. In view of the previous aspects it is the purpose of this paper to analyze the effects of various structural measures and other parameters on the seismic response of such high cooling towers. (orig.)

  3. German seismic regulations

    International Nuclear Information System (INIS)

    Danisch, Ruediger

    2002-01-01

    Rules and regulations for seismic design in Germany cover the following: seismic design of conventional buildings; and seismic design of nuclear facilities. Safety criteria for NPPs, accident guidelines, and guidelines for PWRs as well as safety standards are cited. Safety standards concerned with NPPs seismic design include basic principles, soil analysis, design of building structures, design of mechanical and electrical components, seismic instrumentation, and measures to be undertaken after the earthquake

  4. Design of components of reinforced concrete stressed by seismic loads

    International Nuclear Information System (INIS)

    Sitka, R.

    1980-01-01

    The example of the type of frame investigated shows that the ductility of the system assumed for standard dimensioning of such a frame lies between two and four. According to the system and the loading different requirements may result for the cross-section, that will have to be observed in design. Derived from these requirements rules are given for the design of frames stiffening in horizontal direction that will guarantee a minimum level of ductility. These rules concern the design of joint and node regions, utilization of the compressive force of the concrete as well as guidance and graduation of the reinforcement according to stud and bolt. By means of some examples of damaged components the effects of violating these rules are made clear. (orig./DG) [de

  5. Eccentric bracing of steel frames in seismic design

    International Nuclear Information System (INIS)

    Popov, E.P.; Manheim, D.

    1981-01-01

    The general concepts of designing eccentrically braced steel frames are discussed. A number of possible bracing configurations are pointed out which are suitable for this type of framing. The necessity for considering the collapse mechanism for the selected frame is brought out, and the need for considering the ductility demands for the critical elements is indicated. The need for web stiffness along the critical beam elements (links), and the necessity for lateral bracing at the potential plastic hinges is emphasized. Properly designed eccentrically braced frames provide good drift control for moderate earthquakes, and good ductility for extreme earthquakes. Experience gained in practice attests to the practicality and economy of this kind of framing. The major disadvantage of properly designed eccentrically braced frames lies in the fact that high local distortions may occur during a severe earthquake requiring repair. However, such severe distortions should attenuate rapidly from the damaged areas. (orig./HP)

  6. Development of tools to manage the operational monitoring and pre-design of the NPP-LV cycle; Desarrollo de herramientas para administrar el seguimiento operativo y el pre-diseno del ciclo de la CLV

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Arredondo S, C.; Hernandez M, J. L.; Montes T, J. L.; Castillo M, A.; Ortiz S, J. J., E-mail: raul.perusquia@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    This paper presents the development of tools to facilitate the management so much, the operational monitoring of boiling water reactors (BWR) of the nuclear power plant of Laguna Verde (NPP-LV) through independent codes, and how to carry out the static calculations corresponding to process of optimized pre-design of the reference cycle next to current cycle. The progress and preliminary results obtained with the program SACal, developed at Instituto Nacional de Investigaciones Nucleares (ININ), central tool to achieve provide a management platform of the operational monitoring and pre-design of NPP-LV cycle are also described. The reached preliminary advances directed to get an Analysis center and automated design of fuel assembly cells are also presented, which together with centers or similar modules related with the fuel reloads form the key part to meet the targets set for the realization of a Management Platform of Nuclear Fuel of the NPP-LV. (Author)

  7. Seismic margin analysis technique for nuclear power plant structures

    International Nuclear Information System (INIS)

    Seo, Jeong Moon; Choi, In Kil

    2001-04-01

    In general, the Seismic Probabilistic Risk Assessment (SPRA) and the Seismic Margin Assessment(SAM) are used for the evaluation of realistic seismic capacity of nuclear power plant structures. Seismic PRA is a systematic process to evaluate the seismic safety of nuclear power plant. In our country, SPRA has been used to perform the probabilistic safety assessment for the earthquake event. SMA is a simple and cost effective manner to quantify the seismic margin of individual structural elements. This study was performed to improve the reliability of SMA results and to confirm the assessment procedure. To achieve this goal, review for the current status of the techniques and procedures was performed. Two methodologies, CDFM (Conservative Deterministic Failure Margin) sponsored by NRC and FA (Fragility Analysis) sponsored by EPRI, were developed for the seismic margin review of NPP structures. FA method was originally developed for Seismic PRA. CDFM approach is more amenable to use by experienced design engineers including utility staff design engineers. In this study, detailed review on the procedures of CDFM and FA methodology was performed

  8. Reanalysis and evaluation of seismic response of reactor building

    International Nuclear Information System (INIS)

    Li Zhongcheng; Li Zhongxian

    2005-01-01

    For the Ling Ao phase-I (LA-I) Nuclear Power Plant (NPP), its' seismic analysis of nuclear island was in accordance with the approaches in RCC-G standard for the model M310 in France, in which the Simplified impedance method was employed for the consideration of SSI. Thanks to the rapid progress being made in upgrading the evaluation technology and the capability of data processing systems, methods and software tools for the SSI analysis have experienced significant development all over the world. Focused on the model of reactor building of the LA-I NPP, in this paper the more sophisticated 3D half-space continuum impedance method based on the Green functions is used to analyze the functions of the soil, and then the seismic responses of the coupled SSI system are calculated and compared with the corresponding design values. It demonstrates that the design method provides a set of conservatively safe results. The conclusions from the study are hopefully to provide some important references to the assessment of seismic safety margin for LA-I NPP. (authors)

  9. Effect of URM infills on seismic vulnerability of Indian code designed RC frame buildings

    Science.gov (United States)

    Haldar, Putul; Singh, Yogendra; Paul, D. K.

    2012-03-01

    Unreinforced Masonry (URM) is the most common partitioning material in framed buildings in India and many other countries. Although it is well-known that under lateral loading the behavior and modes of failure of the frame buildings change significantly due to infill-frame interaction, the general design practice is to treat infills as nonstructural elements and their stiffness, strength and interaction with the frame is often ignored, primarily because of difficulties in simulation and lack of modeling guidelines in design codes. The Indian Standard, like many other national codes, does not provide explicit insight into the anticipated performance and associated vulnerability of infilled frames. This paper presents an analytical study on the seismic performance and fragility analysis of Indian code-designed RC frame buildings with and without URM infills. Infills are modeled as diagonal struts as per ASCE 41 guidelines and various modes of failure are considered. HAZUS methodology along with nonlinear static analysis is used to compare the seismic vulnerability of bare and infilled frames. The comparative study suggests that URM infills result in a significant increase in the seismic vulnerability of RC frames and their effect needs to be properly incorporated in design codes.

  10. Some considerations for establishing seismic design criteria for nuclear plant piping

    International Nuclear Information System (INIS)

    Chen, W.P.; Chokshi, N.C.

    1997-01-01

    The Energy Technology Engineering Center (ETEC) is providing assistance to the U.S. NRC in developing regulatory positions on the seismic analysis of piping. As part of this effort, ETEC previously performed reviews of the ASME Code, Section III piping seismic design criteria as revised by the 1994 Addenda. These revised criteria were based on evaluations by the ASME Special Task Group on Integrated Piping Criteria (STGIPC) and the Technical Core Group (TCG) of the Advanced Reactor Corporation (ARC) of the earlier joint Electric Power Research Institute (EPRI)/NRC Piping ampersand Fitting Dynamic Reliability (PFDR) program. Previous ETEC evaluations reported at the 23rd WRSM of seismic margins associated with the revised criteria are reviewed. These evaluations had concluded, in part, that although margins for the timed PFDR tests appeared acceptable (>2), margins in detuned tests could be unacceptable (<1). This conclusion was based primarily on margin reduction factors (MRFs) developed by the ASME STGIPC and ARC/TCG from realistic analyses of PFDR test 36. This paper reports more recent results including: (1) an approach developed for establishing appropriate seismic margins based on PRA considerations, (2) independent assessments of frequency effects on margins, (3) the development of margins based on failure mode considerations, and (4) the implications of Code Section III rules for Section XI

  11. Seismic design of steel moment resisting frames-European versus American practice

    International Nuclear Information System (INIS)

    Naqash, M.T.; Matteis, G.D.; Luca, A.D.

    2012-01-01

    This paper provides an overview on the design philosophy of moment resisting frames (MRF) according to the seismic provisions of Eurocode 8 and American Institute of Steel Construction (AISC). A synopsis of the main recommendations of the two codes is briefly described. Then in order to examine the structural efficiency of the design principles of MRF according to the aforementioned codes, a case study is developed in which spatial and perimeter moment resisting frames of 12, 6 and 3 storeys residential building are considered. In the case of EC8, Ductility Class Medium (DCM) with behaviour factor of 4 and Ductility Class High (DCH) with behaviour factor of 6.5 for 6-storey frames are used, while only DCH is employed in the design of 12 and 3 storey frames. When dealing with AISC/American Society of Civil Engineers (ASCE) code, special moment resisting frame (SMF) with response modification factor of 8 is employed in the design. The outcomes from the design are illustrated in terms of frame performance, section profiles, strength-demand to capacity ratios, drift-demand to capacity ratios and structural weight, thus allowing the understanding of pros and cons of the design criteria and the capacity design rules of the two codes. The main purpose of the current paper is to compare the seismic design rules of the two codes with a parametric analysis developed by a case study in order to let the technician knows about the importance and influence of some important parameters which are given in the capacity design rules of the two codes. This study will be a benchmark for further analysis on the two codes for seismic design of steel structures. (author)

  12. Calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs shake table investigation at Paks NPP. Final report from 15 June 1993 - 15 June 1994

    International Nuclear Information System (INIS)

    Kaznovsky, S.

    1995-01-01

    This final report involves the calculation and experimental examination and ensuring the seismic resistance of the reactor equipment and pipelines at start up and operation of WWER type nuclear power plants. Shake table experiments performed at the Paks NPP are included. Namely the following devices of the emergency cooling system were tested: pump of low pressure; valve of low pressure; intermediate heat exchanger. The following values were determined: natural frequencies and vibration decrements and the main modes of normal vibrations for the heat exchanger

  13. Seismic analysis, support design and stress calculation of HTR-PM transport and conversion devices

    International Nuclear Information System (INIS)

    Zhang Zheyu; Yuan Chaolong; Zhang Haiquan; Nie Junfeng

    2012-01-01

    Background: The transport and conversion devices are important guarantees for normal operation of HTR-PM fuel handling system in normal and fault conditions. Purpose: A conflict of devices' support design needs to be solved. The flexibility of supports is required because of pipe thermal expansion displacement, while the stiffness is also required because of large devices quality and eccentric distance. Methods: In this paper, the numerical simulation was employed to analyze the seismic characteristics and optimize the support program, Under the chosen support program, the stress calculation of platen support bracket was designed by solidworks software. Results: The supports solved the conflict between the flexibility and stiffness requirements. Conclusions: Therefore, it can ensure the safety of transport and conversion devices and the supports in seismic conditions. (authors)

  14. Seismic design technology for breeder reactor structures. Volume 1. Special topics in earthquake ground motion

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

    This report is divided into twelve chapters: seismic hazard analysis procedures, statistical and probabilistic considerations, vertical ground motion characteristics, vertical ground response spectrum shapes, effects of inclined rock strata on site response, correlation of ground response spectra with intensity, intensity attenuation relationships, peak ground acceleration in the very mean field, statistical analysis of response spectral amplitudes, contributions of body and surface waves, evaluation of ground motion characteristics, and design earthquake motions

  15. Preliminary proposed seismic design and evaluation criteria for new and existing underground hazardous materials storage tanks

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1991-01-01

    The document provides a recommended set of deterministic seismic design and evaluation criteria for either new or existing underground hazardous materials storage tanks placed in either the high hazard or moderate hazard usage catagories of UCRL-15910. The criteria given herein are consistent with and follow the same philosophy as those given in UCRL-15910 for the US Department of Energy facilities. This document is intended to supplement and amplify upon Reference 1 for underground hazardous materials storage tanks

  16. Sloped Connections and Connections with Fillet Welded Continuity Plates for Seismic Design of Special Moment Frames

    OpenAIRE

    Mashayekh, Adel

    2017-01-01

    Steel Special Moment Frames (SMF) are one of the most popular lateral force-resisting systems for multistory building construction in high seismic regions due to their architectural versatility. With a significant amount of research that was conducted after the 1994 Northridge, California earthquake, AISC has published design guidelines (AISC 341 and AISC 358) to avoid brittle fracture of beam-to-column welded moment connections that occurred in more than 100 steel buildings. This dissertat...

  17. Improvements of seismic design of nuclear power plant equipment

    International Nuclear Information System (INIS)

    Suzuki, Kohei; Takayama, Yoshihiro.

    1997-01-01

    A brief survey and overview of the current research and development in Japan was presented. Particularly, several kinds of new dampers and isolators were developed and those effectiveness were examined by caring out the large-scale vibration test and so on. The evaluation of the energy absorption of these damping devices at the earthquake appeared to be significant. In addition, it must be necessary to investigate the design margin and the failure mode and limit problem to these devices and the nuclear structures and piping supported by those. Mutual exchange of the information related to these technology and research has to be put forward and cooperative works including the international conference on those issues should be promoted. (J.P.N.)

  18. Improvements of seismic design of nuclear power plant equipment

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Kohei [Tokyo Metropolitan Univ., Hachioji (Japan). Faculty of Technology; Takayama, Yoshihiro

    1997-03-01

    A brief survey and overview of the current research and development in Japan was presented. Particularly, several kinds of new dampers and isolators were developed and those effectiveness were examined by caring out the large-scale vibration test and so on. The evaluation of the energy absorption of these damping devices at the earthquake appeared to be significant. In addition, it must be necessary to investigate the design margin and the failure mode and limit problem to these devices and the nuclear structures and piping supported by those. Mutual exchange of the information related to these technology and research has to be put forward and cooperative works including the international conference on those issues should be promoted. (J.P.N.)

  19. Modeling of the fatigue damage accumulation processes in the material of NPP design units under thermomechanical unstationary effects. Estimation of spent life and forecast of residual life

    International Nuclear Information System (INIS)

    Kiriushin, A.I.; Korotkikh, Yu.G.; Gorodov, G.F.

    2002-01-01

    Full text: The estimation problems of spent life and forecast of residual life of NPP equipment design units, operated at unstationary thermal force loads are considered. These loads are, as a rule, unregular and cause rotation of main stress tensor platforms of the most loaded zones of structural elements and viscoelastic plastic deformation of material in the places of stresses concentrations. The existing engineering approaches to the damages accumulation processes calculation in the material of structural units, their advantages and disadvantages are analyzed. For the processes of fatigue damages accumulation a model is proposed, which allows to take into account the unregular pattern of deformation multiaxiality of stressed state, rotation of main platforms, non-linear summation of damages at the loading mode change. The model in based on the equations of damaged medium mechanics, including the equations of viscoplastic deformation of the material and evolutionary equations of damages accumulation. The algorithms of spent life estimation and residual life forecast of the controlled equipment and systems zones are made on the bases of the given model by the known real history of loading, which is determined by real model of NPP operation. The results of numerical experiments on the basis of given model for various processes of thermal force loads and their comparison with experimental results are presented. (author)

  20. Study of seismic design bases for nuclear power plants in the US

    International Nuclear Information System (INIS)

    Kintzer, F.C.; Yanev, P.I.; Gotschall, H.L.

    1983-01-01

    This paper presents the results of an investigation of topics pertinent to establishing design basis seismic events and soil conditions for deployment of the High Temperature Gas-Cooled Reactor - Steam Cycle/Cogeneration (HTGR-SC/C) System. Generalized design ground accelerations and soil shear wave velocities are presented by regions of the continental United States. Design basis accelerations and soil conditions for existing nuclear power plants are summarized. Finally, analytical approaches to assess soil-structure interaction, including the effects of embedment, are reviewed

  1. Seismic Performance and Design of Steel Plate Shear Walls with Low Yield Point Steel Infill Plates

    OpenAIRE

    Zirakian, Tadeh

    2013-01-01

    Steel plate shear walls (SPSWs) have been frequently used as the primary or part of the primary lateral force-resisting system in design of low-, medium-, and high-rise buildings. Their application has been based on two different design philosophies as well as detailing strategies. Stiffened and/or stocky-web SPSWs with improved buckling stability and high seismic performance have been mostly used in Japan, which is one of the pioneering countries in design and application of these systems. U...

  2. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  3. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

    2014-01-01

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  4. Seismic design criteria used for electrical raceway systems in commercial nuclear power plants

    International Nuclear Information System (INIS)

    Summers, P.B.; Manrique, M.A.; Nelson, T.A.

    1991-01-01

    This paper summarizes some of the seismic design approaches, relevant technical issues and criteria used over the years for design of electrical raceway systems at commercial nuclear power plant facilities. The approaches used for design and endorsed by the NRC can be seen to be quite varied. In recent years, considerably more rigor has been required for raceway design, as well as for the level of design basis documentation produced. However, there has also been a willingness by the NRC to accept rational approaches based on testing, analytical results or experience data, provided proper justification is given. Such rational approaches can simplify the significant task of analysis, design and construction of miles of raceways and thousands of raceway supports. Summarizing past practice and identifying relevant technical issues are an important first step in formalizing up-to-date criteria for new raceway designs

  5. Life-cycle cost assessment of optimally designed reinforced concrete buildings under seismic actions

    International Nuclear Information System (INIS)

    Mitropoulou, Chara Ch.; Lagaros, Nikos D.; Papadrakakis, Manolis

    2011-01-01

    Life-cycle cost analysis (LCCA) is an assessment tool for studying the performance of systems in many fields of engineering. In earthquake engineering LCCA demands the calculation of the cost components that are related to the performance of the structure in multiple earthquake hazard levels. Incremental static and dynamic analyses are two procedures that can be used for estimating the seismic capacity of a structural system and can therefore be incorporated into the LCCA methodology. In this work the effect of the analysis procedure, the number of seismic records imposed, the performance criterion used and the structural type (regular or irregular) is investigated, on the life-cycle cost analysis of 3D reinforced concrete structures. Furthermore, the influence of uncertainties on the seismic response of structural systems and their impact on LCCA is examined. The uncertainty on the material properties, the cross-section dimensions and the record-incident angle is taking into account with the incorporation of the Latin hypercube sampling method into the incremental dynamic analysis procedure. In addition, the LCCA methodology is used as an assessment tool for the designs obtained by means of prescriptive and performance-based optimum design methodologies. The first one is obtained from a single-objective optimization problem, where the initial construction cost was the objective to be minimized, while the second one as a two-objective optimization problem where the life-cycle cost was the additional objective also to be minimized.

  6. Design considerations associated with the response of seismic isolators and real scale energy absorbers

    International Nuclear Information System (INIS)

    Benzoni, Gianmario

    2015-01-01

    Few observations obtained from extensive experimental programs for the characterization of anti-seismic devices are proposed hereafter. Specifically, few current code requirements, originally intended for the acquisition of fundamental characteristics of performance, proved difficult to be implemented and of questionable significance for the design phase of a seismic isolation application. In particular, for commonly used devices as elastomeric and friction-based isolators, the experimentally validated variation of performance parameters is often not addressed in existing codes and typically neglected in structural models, based on extreme simplification of the device behaviour. The goal of this paper is to suggest an update to specific codes but particularly to solicit the designer’s awareness against oversimplification in the modelling phase of the device performance [it

  7. Decision making with epistemic uncertainty under safety constraints: An application to seismic design

    Science.gov (United States)

    Veneziano, D.; Agarwal, A.; Karaca, E.

    2009-01-01

    The problem of accounting for epistemic uncertainty in risk management decisions is conceptually straightforward, but is riddled with practical difficulties. Simple approximations are often used whereby future variations in epistemic uncertainty are ignored or worst-case scenarios are postulated. These strategies tend to produce sub-optimal decisions. We develop a general framework based on Bayesian decision theory and exemplify it for the case of seismic design of buildings. When temporal fluctuations of the epistemic uncertainties and regulatory safety constraints are included, the optimal level of seismic protection exceeds the normative level at the time of construction. Optimal Bayesian decisions do not depend on the aleatory or epistemic nature of the uncertainties, but only on the total (epistemic plus aleatory) uncertainty and how that total uncertainty varies randomly during the lifetime of the project. ?? 2009 Elsevier Ltd. All rights reserved.

  8. SEISMIC DESIGN REQUIREMENTS SELECTION METHODOLOGY FOR THE SLUDGE TREATMENT and M-91 SOLID WASTE PROCESSING FACILITIES PROJECTS

    International Nuclear Information System (INIS)

    RYAN GW

    2008-01-01

    In complying with direction from the U.S. Department of Energy (DOE), Richland Operations Office (RL) (07-KBC-0055, 'Direction Associated with Implementation of DOE-STD-1189 for the Sludge Treatment Project,' and 08-SED-0063, 'RL Action on the Safety Design Strategy (SDS) for Obtaining Additional Solid Waste Processing Capabilities (M-91 Project) and Use of Draft DOE-STD-I 189-YR'), it has been determined that the seismic design requirements currently in the Project Hanford Management Contract (PHMC) will be modified by DOE-STD-1189, Integration of Safety into the Design Process (March 2007 draft), for these two key PHMC projects. Seismic design requirements for other PHMC facilities and projects will remain unchanged. Considering the current early Critical Decision (CD) phases of both the Sludge Treatment Project (STP) and the Solid Waste Processing Facilities (M-91) Project and a strong intent to avoid potentially costly re-work of both engineering and nuclear safety analyses, this document describes how Fluor Hanford, Inc. (FH) will maintain compliance with the PHMC by considering both the current seismic standards referenced by DOE 0 420.1 B, Facility Safety, and draft DOE-STD-1189 (i.e., ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, and ANSI ANS 2.26-2004, Categorization of Nuclear Facility Structures, Systems and Components for Seismic Design, as modified by draft DOE-STD-1189) to choose the criteria that will result in the most conservative seismic design categorization and engineering design. Following the process described in this document will result in a conservative seismic design categorization and design products. This approach is expected to resolve discrepancies between the existing and new requirements and reduce the risk that project designs and analyses will require revision when the draft DOE-STD-1189 is finalized

  9. Optimum Performance-Based Seismic Design Using a Hybrid Optimization Algorithm

    Directory of Open Access Journals (Sweden)

    S. Talatahari

    2014-01-01

    Full Text Available A hybrid optimization method is presented to optimum seismic design of steel frames considering four performance levels. These performance levels are considered to determine the optimum design of structures to reduce the structural cost. A pushover analysis of steel building frameworks subject to equivalent-static earthquake loading is utilized. The algorithm is based on the concepts of the charged system search in which each agent is affected by local and global best positions stored in the charged memory considering the governing laws of electrical physics. Comparison of the results of the hybrid algorithm with those of other metaheuristic algorithms shows the efficiency of the hybrid algorithm.

  10. Decommissioning of Brennilis NPP

    International Nuclear Information System (INIS)

    Baize, Jean-Marc

    1998-01-01

    This EDF press communique give information related to the decommissioning of the Brennilis NPP. The following five items are developed in this report: 1. the level-2 decommissioning operations at the Brennilis NPP; 2. the Brennilis NPP, a pilot operation from the commissioning up to the decommissioning; 3. history of the Brennilis NPP decommissioning; 4. the types of radioactive wastes generated by the Brennilis NPP decommissioning; 5. the Brennilis NPP - a yard management as a function of the wastes. The document contains also seven appendices addressing the following subjects: 1. the share of decommissioning assigned to EDF and the decommissioning steps; 2. the EDF installations in course of decommissioning; 3. the CEA decommissioned installations or in course of decommissioning; 4. regulations; 5. costs; 6. waste management - principles; 7. data on the decommissioning yard

  11. NPP life management (abstracts)

    International Nuclear Information System (INIS)

    Litvinskij, L.L.; Barbashev, S.V.

    2002-01-01

    Abstracts of the papers presented at the International conference of the Ukrainian Nuclear Society 'NPP Life Management'. The following problems are considered: modernization of the NPP; NPP life management; waste and spent nuclear fuel management; decommissioning issues; control systems (including radiation and ecological control systems); information and control systems; legal and regulatory framework. State nuclear regulatory control; PR in nuclear power; training of personnel; economics of nuclear power engineering

  12. Seismic Retrofit of Reinforced Concrete Frame Buildings with Hysteretic Bracing Systems: Design Procedure and Behaviour Factor

    Directory of Open Access Journals (Sweden)

    Antonio Di Cesare

    2017-01-01

    Full Text Available This paper presents a design procedure to evaluate the mechanical characteristics of hysteretic Energy Dissipation Bracing (EDB systems for seismic retrofitting of existing reinforced concrete framed buildings. The proposed procedure, aiming at controlling the maximum interstorey drifts, imposes a maximum top displacement as function of the seismic demand and, if needed, regularizes the stiffness and strength of the building along its elevation. In order to explain the application of the proposed procedure and its capacity to involve most of the devices in the energy dissipation with similar level of ductility demand, a simple benchmark structure has been studied and nonlinear dynamic analyses have been performed. A further goal of this work is to propose a simplified approach for designing dissipating systems based on linear analysis with the application of a suitable behaviour factor, in order to achieve a widespread adoption of the passive control techniques. At this goal, the increasing of the structural performances due to the addition of an EDB system designed with the above-mentioned procedure has been estimated considering one thousand case studies designed with different combinations of the main design parameters. An analytical formulation of the behaviour factor for braced buildings has been proposed.

  13. Mechanical design of a single-axis monolithic accelerometer for advanced seismic attenuation systems

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, Alessandro [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy) and LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States)]. E-mail: alessandro.bertolini@desy.de; DeSalvo, Riccardo [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Fidecaro, Francesco [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Francesconi, Mario [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Marka, Szabolcs [Department of Physics, Columbia University, 538 W. 120th St., New York, NY 10027 (United States); Sannibale, Virginio [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Simonetti, Duccio [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Takamori, Akiteru [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Earthquake Research Institute, University of Tokyo, 1-1-1 Yayoi, Bunkyo-Ku, Tokyo 113-0032 (Japan); Tariq, Hareem [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States)

    2006-01-15

    The design and mechanics for a new very-low noise low frequency horizontal accelerometer is presented. The sensor has been designed to be integrated in an advanced seismic isolation system for interferometric gravitational wave detectors. The motion of a small monolithic folded-pendulum (FP) is monitored by a high resolution capacitance displacement sensor; a feedback force actuator keeps the mass at the equilibrium position. The feedback signal is proportional to the ground acceleration in the frequency range 0-150Hz. The very high mechanical quality factor, Q{approx}3000 at a resonant frequency of 0.5Hz, reduces the Brownian motion of the proof mass of the accelerometer below the resolution of the displacement sensor. This scheme enables the accelerometer to detect the inertial displacement of a platform with a root-mean-square noise less than 1nm, integrated over the frequency band from 0.01 to 150Hz. The FP geometry, combined with the monolithic design, allows the accelerometer to be extremely directional. A vertical-horizontal coupling ranging better than 10{sup -3} has been achieved. A detailed account of the design and construction of the accelerometer is reported here. The instrument is fully ultra-high vacuum compatible and has been tested and approved for integration in seismic attenuation system of japanese TAMA 300 gravitational wave detector. The monolithic design also makes the accelerometer suitable for cryogenic operation.

  14. Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

    International Nuclear Information System (INIS)

    Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

    1995-01-01

    This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

  15. Seismic PSA implementation standards by AESJ and the utilization of the advanced safety examination guideline for seismic design for nuclear power plant

    International Nuclear Information System (INIS)

    Ebisawa, Katsumi; Hibino, Kenta

    2008-01-01

    The Advanced Safety Examination Guideline for Seismic Design for Nuclear Power Plant (the advanced safety examination guideline) was worked out on September 19, 2006. In this paper, a summary of the method of probability theory in the advanced safety examination guideline and the Seismic PSA Implementation Standards is stated. On utilization of the probability theory for the advanced safety examination guideline, the uncertainty resulting from the process of the decision of the basic design earthquake ground motion (Ss) is stated to be considered using the proper method. The references of the extra probability for evaluation of earthquake hazard and combination of the working load and the earthquake load are stated. Definition, evaluation method and effort to lower the 'residual risks', and relation between the residual risks and the extra probability of Ss are described. A summary of the earthquake-resistant design for nuclear power facilities is explained by the old guideline. (S.Y.)

  16. Treatment of NPP wastes using vitrification

    International Nuclear Information System (INIS)

    Sobolev, I.A.; Lifanov, F.A.; Stefanovsky, S.V.; Kobelev, A.P.; Savkin, A.E.; Kornev, V.I.

    1998-01-01

    Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10 -5 -10 -6 g/(cm 2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

  17. Harmonizing seismic hazard assessments for nuclear power plants

    International Nuclear Information System (INIS)

    Mallard, D.J.

    1993-01-01

    Even a cursory comparison between maps of global seismicity and NPP earthquake design levels reveals many inconsistencies. While, in part, this situation reflects the evolution in understanding of seismic hazards, mismatches can also be due to ongoing differences in the way the hazards are assessed and in local regulatory requirements. So far, formal international consensus has only been able to encompass broad principles, such as those recently recommended by the International Atomic Energy Agency, and even these can raise many technical issues, particularly relating to zones of diffuse seismicity. In the future, greater harmonisation in hazard assessments and, to some extent, in earthquake design levels could emerge through the more widespread use of probabilistic methods. International collaborative ventures and joint projects will be important for resolving anomalies in the existing databases and their interpretations, and for acquiring new data, but to achieve their ideal objectives, they will need to proceed in clearly defined stages. (author)

  18. Disaggregated seismic hazard and the elastic input energy spectrum: An approach to design earthquake selection

    Science.gov (United States)

    Chapman, Martin Colby

    1998-12-01

    The design earthquake selection problem is fundamentally probabilistic. Disaggregation of a probabilistic model of the seismic hazard offers a rational and objective approach that can identify the most likely earthquake scenario(s) contributing to hazard. An ensemble of time series can be selected on the basis of the modal earthquakes derived from the disaggregation. This gives a useful time-domain realization of the seismic hazard, to the extent that a single motion parameter captures the important time-domain characteristics. A possible limitation to this approach arises because most currently available motion prediction models for peak ground motion or oscillator response are essentially independent of duration, and modal events derived using the peak motions for the analysis may not represent the optimal characterization of the hazard. The elastic input energy spectrum is an alternative to the elastic response spectrum for these types of analyses. The input energy combines the elements of amplitude and duration into a single parameter description of the ground motion that can be readily incorporated into standard probabilistic seismic hazard analysis methodology. This use of the elastic input energy spectrum is examined. Regression analysis is performed using strong motion data from Western North America and consistent data processing procedures for both the absolute input energy equivalent velocity, (Vsbea), and the elastic pseudo-relative velocity response (PSV) in the frequency range 0.5 to 10 Hz. The results show that the two parameters can be successfully fit with identical functional forms. The dependence of Vsbea and PSV upon (NEHRP) site classification is virtually identical. The variance of Vsbea is uniformly less than that of PSV, indicating that Vsbea can be predicted with slightly less uncertainty as a function of magnitude, distance and site classification. The effects of site class are important at frequencies less than a few Hertz. The regression

  19. SRS BEDROCK PROBABILISTIC SEISMIC HAZARD ANALYSIS (PSHA) DESIGN BASIS JUSTIFICATION (U)

    Energy Technology Data Exchange (ETDEWEB)

    (NOEMAIL), R

    2005-12-14

    This represents an assessment of the available Savannah River Site (SRS) hard-rock probabilistic seismic hazard assessments (PSHAs), including PSHAs recently completed, for incorporation in the SRS seismic hazard update. The prior assessment of the SRS seismic design basis (WSRC, 1997) incorporated the results from two PSHAs that were published in 1988 and 1993. Because of the vintage of these studies, an assessment is necessary to establish the value of these PSHAs considering more recently collected data affecting seismic hazards and the availability of more recent PSHAs. This task is consistent with the Department of Energy (DOE) order, DOE O 420.1B and DOE guidance document DOE G 420.1-2. Following DOE guidance, the National Map Hazard was reviewed and incorporated in this assessment. In addition to the National Map hazard, alternative ground motion attenuation models (GMAMs) are used with the National Map source model to produce alternate hazard assessments for the SRS. These hazard assessments are the basis for the updated hard-rock hazard recommendation made in this report. The development and comparison of hazard based on the National Map models and PSHAs completed using alternate GMAMs provides increased confidence in this hazard recommendation. The alternate GMAMs are the EPRI (2004), USGS (2002) and a regional specific model (Silva et al., 2004). Weights of 0.6, 0.3 and 0.1 are recommended for EPRI (2004), USGS (2002) and Silva et al. (2004) respectively. This weighting gives cluster weights of .39, .29, .15, .17 for the 1-corner, 2-corner, hybrid, and Greens-function models, respectively. This assessment is judged to be conservative as compared to WSRC (1997) and incorporates the range of prevailing expert opinion pertinent to the development of seismic hazard at the SRS. The corresponding SRS hard-rock uniform hazard spectra are greater than the design spectra developed in WSRC (1997) that were based on the LLNL (1993) and EPRI (1988) PSHAs. The

  20. Assessing seismic adequacy of existing nuclear power plant structures

    International Nuclear Information System (INIS)

    Belyaev, V.; Vinogradov, V.; Privalov, S.; Shishenin, V.

    2003-01-01

    Nowadays Russia's specialists perform a huge amount of works to revaluate the NPP safety. These works are certain to include refinement of NPP safety assessment under the effects of specific dynamic loads, earthquake effects included. It should be noted, that a number of Russian NPPs now in operation had been designed either with no account of these loads, or under the requirements which are underestimated as compared with the modern requirements on the external load composition and rate. Revaluation of NPP seismic safety is based on the results of the works taken under orderly sequence on assessment of (1) seismic input and ground effects; (2) structure response and state; (3) equipment and pipelines response and state. The paper considers the methods of NPP structures response and state assessment. Therewith we assume that ground motion predicted behavior at the construction basement has been preset for the SSE and OBE conditions and the effects of soil-structure interaction, including the situation of possible soft soil liquefaction. Necessity to determine both the reaction of a construction and its state as a whole as well as its elements reaction, to evaluate their bearing capacity and destruction zones formation makes it necessary to make up a detailed structural model, which is usually a finite element one. Since seismic revaluation is to be performed for the existing structures, characteristics of which can substantially differ from the design ones, revealing the actual state of this structures becomes critical. If the real values of physical and mechanical properties of the structure materials, connections of elements etc. are used as initial data in a structural model this permits to increase the design assessment credibility and reliability substantially. The paper analyzes the results of determining these initial assessments while inspecting several Russian NPPs on the basis of a 'combined' method. This method is realized at two consecutive stages. The

  1. Generation of artificial earthquake time histories for seismic design at Hanford, Washington

    International Nuclear Information System (INIS)

    Salmon, M.W.; Kuilanoff, G.

    1991-01-01

    The purpose of the development of artificial time-histories is to provide the designer with ground motion estimates which will meet the requirements of the design guidelines at the Hanford site. In particular, the artificial time histories presented in this paper were prepared to assist designers of the Hanford Waste Vitrification Plant (HWVP) with time histories that envelop the requirements for both a large magnitude earthquake (MI > 6.0) and a small magnitude, near-field earthquake (MI < 5. 0). A background of the requirements for both the large magnitude and small magnitude events is presented in this paper. The work done in generating time histories which produce response spectra matching those of the design seismic events is also presented. Finally, some preliminary results from studies performed using the small-magnitude near-filed earthquake time-history are presented

  2. A recommended epidemiological study design for examining the adverse health effects among emergency workers who experienced the TEPCO fukushima daiichi NPP accident in 2011.

    Science.gov (United States)

    Yasui, Shojiro

    2016-01-01

    Results from medical examinations conducted in 2012 of workers who were engaged in radiation work in 2012 as a result of the 2011 Fukushima Daiichi Nuclear Power Plant (NPP) accident showed that the prevalence of abnormal findings was 4.21%, 3.23 points higher than the 0.98% that was found prior to the accident in the jurisdiction area of the labor inspection office which holds jurisdiction over the NPP. The Ministry of Health, Labour and Welfare (MHLW) concluded that the 2010 and 2012 data cannot be easily compared because 70% of the enterprises within the jurisdiction of the office that reported the 2012 results were different from those that did so in 2010. In addition, although the radiation workers' estimated average dose weighted by number of workers was 3.66 times higher than decontamination workers' dose, the prevalence among radiation workers was only 1.14 times higher than that among decontamination workers. Based on the results of the medical examinations, however, the MHLW decided to implement an epidemiological study on the health effects of radiation exposure on all emergency workers. This article explains key issues of the basic design of the study recommended by the expert meeting established in the MHLW and also identifies challenges that could not be resolved and thus required further consideration by the study researchers. The major issues included: (a) study methods and target group; (b) evaluation of cumulative doses; (c) health effects (end points); (d) control of confounding factors; and (e) study implementation framework. Identified key challenges that required further deliberation were: (a) preventing arbitrary partisan analysis; (b) ensuring a high participation rate; (c) inquiry about the medical radiation doses; and (d) the preparedness of new analytical technology. The study team formulated and implemented the pilot study in 2014 and started the full-scale study in April 2015 with funding from a research grant from the MHLW.

  3. Recommended revisions to Nuclear Regulatory Commission seismic design criteria. Technical report

    International Nuclear Information System (INIS)

    Coats, D.W.

    1980-05-01

    This report recommends changes in the Nuclear Regulatory Commission's (NRC's) criteria now used in the seismic design of nuclear power plants. Areas covered include ground motion, soil-structure interaction, structures, and equipment and components. Members of the Engineering Mechanics Section of the Nuclear Test Engineering Division at Lawrence Livermore Laboratory (LLL) generally agreed upon the recommendations, which are based on (1) reports developed under the NRC's Task Action Plan A-40, (2) other available engineering literature, and (3) recommendations of nationally recognized experts retained by LLL specifically for this task

  4. Design and realization of real-time processing system for seismic exploration

    International Nuclear Information System (INIS)

    Zhang Sifeng; Cao Ping; Song Kezhu; Yao Lin

    2010-01-01

    For solving real-time seismic data processing problems, a high-speed, large-capacity and real-time data processing system is designed based on FPGA and ARM. With the advantages of multi-processor, DRPS has the characteristics of high-speed data receiving, large-capacity data storage, protocol analysis, data splicing, data converting from time sequence into channel sequence, no dead time data ping-pong storage, etc. And with the embedded Linux operating system, DRPS has the characteristics of flexibility and reliability. (authors)

  5. Evaluation of seismic design by students made after Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    Sugiyama, Ken-ichiro

    2012-01-01

    The sense of anxiety for safety of nuclear power plants among people in Japan has not disappeared after Fukushima Dai-ichi accident because of a typical country with frequent earthquakes. The provision of information for seismic design in nuclear power plants prepared for easier comprehension is always required in any kind of study meetings for the social acceptance of nuclear power plants. In the present paper, the effect of the provision of information made an attempt for students in Hokkaido University is reported. (author)

  6. Seismic methodology in determining basis earthquake for nuclear installation

    International Nuclear Information System (INIS)

    Ameli Zamani, Sh.

    2008-01-01

    Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

  7. Seismic Ecology

    Science.gov (United States)

    Seleznev, V. S.; Soloviev, V. M.; Emanov, A. F.

    The paper is devoted to researches of influence of seismic actions for industrial and civil buildings and people. The seismic actions bring influence directly on the people (vibration actions, force shocks at earthquakes) or indirectly through various build- ings and the constructions and can be strong (be felt by people) and weak (be fixed by sensing devices). The great number of work is devoted to influence of violent seismic actions (first of all of earthquakes) on people and various constructions. This work is devoted to study weak, but long seismic actions on various buildings and people. There is a need to take into account seismic oscillations, acting on the territory, at construction of various buildings on urbanized territories. Essential influence, except for violent earthquakes, man-caused seismic actions: the explosions, seismic noise, emitted by plant facilities and moving transport, radiation from high-rise buildings and constructions under action of a wind, etc. can exert. Materials on increase of man- caused seismicity in a number of regions in Russia, which earlier were not seismic, are presented in the paper. Along with maps of seismic microzoning maps to be built indicating a variation of amplitude spectra of seismic noise within day, months, years. The presence of an information about amplitudes and frequencies of oscillations from possible earthquakes and man-caused oscillations in concrete regions allows carry- ing out soundly designing and construction of industrial and civil housing projects. The construction of buildings even in not seismically dangerous regions, which have one from resonance frequencies coincident on magnitude to frequency of oscillations, emitted in this place by man-caused objects, can end in failure of these buildings and heaviest consequences for the people. The practical examples of detail of engineering- seismological investigation of large industrial and civil housing projects of Siberia territory (hydro power

  8. Improved Simplified Methods for Effective Seismic Analysis and Design of Isolated and Damped Bridges in Western and Eastern North America

    Science.gov (United States)

    Koval, Viacheslav

    The seismic design provisions of the CSA-S6 Canadian Highway Bridge Design Code and the AASHTO LRFD Seismic Bridge Design Specifications have been developed primarily based on historical earthquake events that have occurred along the west coast of North America. For the design of seismic isolation systems, these codes include simplified analysis and design methods. The appropriateness and range of application of these methods are investigated through extensive parametric nonlinear time history analyses in this thesis. It was found that there is a need to adjust existing design guidelines to better capture the expected nonlinear response of isolated bridges. For isolated bridges located in eastern North America, new damping coefficients are proposed. The applicability limits of the code-based simplified methods have been redefined to ensure that the modified method will lead to conservative results and that a wider range of seismically isolated bridges can be covered by this method. The possibility of further improving current simplified code methods was also examined. By transforming the quantity of allocated energy into a displacement contribution, an idealized analytical solution is proposed as a new simplified design method. This method realistically reflects the effects of ground-motion and system design parameters, including the effects of a drifted oscillation center. The proposed method is therefore more appropriate than current existing simplified methods and can be applicable to isolation systems exhibiting a wider range of properties. A multi-level-hazard performance matrix has been adopted by different seismic provisions worldwide and will be incorporated into the new edition of the Canadian CSA-S6-14 Bridge Design code. However, the combined effect and optimal use of isolation and supplemental damping devices in bridges have not been fully exploited yet to achieve enhanced performance under different levels of seismic hazard. A novel Dual-Level Seismic

  9. Basic concepts about application of dual vibration absorbers to seismic design of nuclear piping systems

    International Nuclear Information System (INIS)

    Hara, F.; Seto, K.

    1987-01-01

    The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers

  10. Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1993-06-01

    In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP is comprised of a number of consultants known for their knowledge of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation Guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects which might employ the TSEP guidelines

  11. Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1993-01-01

    In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the U.S. Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP comprises a number of consultants known for their understanding of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects that might employ the TSEP guidelines

  12. Research program for seismic qualification of nuclear plant electrical and mechanical equipment. Task 4. Use of fragility in seismic design of nuclear plant equipment. Volume 4

    International Nuclear Information System (INIS)

    Kana, D.D.; Pomerening, D.J.

    1984-08-01

    The Research Program for Seismic Qualification of Nuclear Plant Electrical and Mechanical Equipment has spanned a period of three years and resulted in seven technical summary reports, each of which have covered in detail the findings of different tasks and subtasks, and have been combined into five NUREG/CR volumes. Volume 4 presents study of the use of fragility concepts in the design of nuclear plant equipment and compares the results of state-of-the-art proof testing with fragility testing

  13. Nuclear Fuel in Cofrentes NPP

    International Nuclear Information System (INIS)

    2002-01-01

    Fuel is an essential in the nuclear power generating business because of its direct implications on safety, generating costs and the operating conditions and limitations of the facility. Fuel management in Cofrentes NPP has been targeted at optimized operation, enhanced reliability and the search for an in-depth knowledge of the design and licensing processes that will provide Iberdrola,as the responsible operator, with access to independent control of safety aspects related to fuel and free access to manufacturing markets. (Author)

  14. Ageing management database development for PWR NPP steam generator

    International Nuclear Information System (INIS)

    Liu Hongyun; Xu Liangjun; Xiong Changhuai; Wang Xianyuan

    2005-01-01

    Steam generator (SG) is one of the key safe important equipment of NPP, which is covered by NPP aging management program. Steam Generator Aging Management Dabatase (SGAMDB) is developed to provide necessary information for SG aging management. RINPO is developing SGAMDB for domestic NPP. This system contains information and data about SG design, manufacture, operation and maintenance. The information include NPP fundamental data, SG design data, SG aging mechanism, SG operation data, SG ISI data, SG maintenance data and SG evaluation interface. The system runs at the intranet of Qinshan-1 NPP with B/S mode. It can provide information inquire and fundamental analysis for NPP SG aging team and SG aging researcher's. In addition, it provides necessary information and data for SG aging analysis and evaluation, such as all pressure test process and flaws of tubes, and collects the analysis results. (authors)

  15. NPP Krsko Living PSA Concept

    International Nuclear Information System (INIS)

    Vrbanic, I.; Spiler, J.

    2000-01-01

    NPP Krsko developed PSA model of internal and external initiators within the frame of the Individual Plant Examination (IPE) project. Within this project PSA model was used to examine the existing plant design features. In order to continue with use of this PSA model upon the completion of IPE in various risk-informed applications in support of plant operation and evaluations of design changes, an appropriate living PSA concept needed to be defined. The Living PSA concept is in NPP Krsko considered as being a set of activities pursued in order to update existing PSA model in a manner that it appropriately represents the plant design, operation practice and history. Only a PSA model which is being updated in this manner can serve as a platform for plant-specific risk informed applications. The NPP Krsko living PSA concept is based on the following major ponts. First, the baseline PSA model is defined, which is to be maintained and updated and which is to be reference point for any risk-informed application. Second, issues having a potential for impact on baseline PSA model are identified and procedure and responsibilities for their permanent monitoring and evaluation are established. Third, manner is defined in which consequential changes to baseline PSA model are implemented and controlled, together with associated responsibilities. Finally, the process is defined by which the existing version of baseline PSA model is superseded by a new one. Each time a new version of baseline PSA model is released, it would be re-quantified and the results evaluated and interpreted. By documenting these re-quantifications and evaluations of results in a sequence, the track is being kept of changes in long-term averaged risk perspective, represented by long-term averaged frequencies of core damage and pre-defined release categories. These major topics of NPP Krsko living PSA concept are presented and discussed in the paper. (author)

  16. Status for seismic design requirements of nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Takahashi, H.

    1977-01-01

    The fundamental purpose for the aseismic design of the nuclear power plants is to protect the inhabitants near the plant from radiation accidents during and after earthquake vibrations. In order to achieve the above purpose, the following considerations have been made. All buidlings, structures, system and components are classified into three Classes A, B and C according to their degree of importance for plant safety, and are designed to meet the requirements specified for each class, respectively. Magnitude and epicenter of the design basis earthquake are determined based upon seismological and geological investigations and observation of ground motion in the site, and the maximum ground acceleration which could be expected can be calculated empirically. With respect to time history waves, more than three are selected referring to dynamic characteristic of base rock in the site, observed ground motion records in the site or other strong motion seismographs.The figures of horizontal seismic coefficients to be used in determining design forces on Class A buildings and structures are 3 Co (where Co. is as defined in the Japan Building Standard Law). On the other hand the horizontal design force should not be less than those determined as the results of the dynamic analyses based on DEGM (Design Earthquake Ground Motion). The figures of horizontal seismic coefficient and forces for Class A system and components are usually determined based on the dynamic analyses for DEGM. The buildings and structures treated as an elastic column system with masses, and the bottom mass is supported by elastic springs representing the soil-foundation interaction characteristics. DEGM is used as the input disturbance in the dynamic response analysis, and the model analysis or time history method is worked out. System and components are modeled as elastic bars with lumped masses of 3 dimensional degree of freedom, and the response analysis is carried out using floor respone spectra

  17. Energy-Based Design Criterion of Dissipative Bracing Systems for the Seismic Retrofit of Frame Structures

    Directory of Open Access Journals (Sweden)

    Gloria Terenzi

    2018-02-01

    Full Text Available Direct sizing criteria represent useful tools in the design of dissipative bracing systems for the advanced seismic protection of existing frame structures, especially when incorporated dampers feature a markedly non-linear behaviour. An energy-based procedure is proposed herein to this aim, focusing attention on systems including fluid viscous devices. The procedure starts by assuming prefixed reduction factors of the most critical response parameters in current conditions, which are evaluated by means of a conventional elastic finite element analysis. Simple formulas relating the reduction factors to the equivalent viscous damping ratio of the dampers, ξeq, are proposed. These formulas allow calculating the ξeq values that guarantee the achievement of the target factors. Finally, the energy dissipation capacity of the devices is deduced from ξeq, finalizing their sizing process. A detailed description of the procedure is presented in the article, by distinguishing the cases where the prevailing structural deficiencies are represented by poor strength of the constituting members, from the cases having excessive horizontal displacements. A demonstrative application to the retrofit design of a reinforced concrete gym building is then offered to explicate the steps of the sizing criterion in practice, as well as to evaluate the enhancement of the seismic response capacities generated by the installation of the dissipative system.

  18. Transition cycle fuel management problems of NPP Krsko

    International Nuclear Information System (INIS)

    Petrovic, B.; Pevec, D.; Smuc, T.; Urli, N.

    1989-01-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  19. Current status of ground motions evaluation in seismic design guide for nuclear power facilities. Investigation on IAEA and US.NRC

    International Nuclear Information System (INIS)

    Nakajima, Masato; Ito, Hiroshi; Hirata, Kazuta

    2009-01-01

    Recently, IAEA (International Atomic Energy Agency) and US.NRC (US. Nuclear Regulatory Commission) published several standards and technical reports on seismic design and safety evaluation for nuclear power facilities. This report summarizes the current status of the international guidelines on seismic design and safety evaluation for nuclear power facilities in order to explore the future research topics. The main results obtained are as follows: 1 IAEA: (1) In the safety standard series, two levels are defined as seismic design levels, and design earthquake ground motion is determined corresponding to each seismic design level. (2) A new framework on seismic design which consists of conventional deterministic method and risk-based method is discussed in the technical report although the framework is not adopted in the safety guidelines. 2 USA: (1) US.NRC discusses a performance-based seismic design framework which has been originally developed by the private organization (American Society of Civil Engineers). (2) Design earthquakes and earthquake ground motion are mainly evaluated and determined based on probabilistic seismic hazard evaluations. 3 Future works: It should be emphasized that IAEA and US.NRC have investigated the implementation of risk-based concept into seismic design. The implementation of risk-based concept into regulation and seismic design makes it possible to consider various uncertainties and to improve accountability. Therefore, we need to develop the methods for evaluating seismic risk of structures, and to correlate seismic margin and seismic risk quantitatively. Moreover, the probabilistic method of earthquake ground motions, that is required in the risk-based design, should be applied to sites in Japan. (author)

  20. AP1000R design robustness against extreme external events - Seismic, flooding, and aircraft crash

    International Nuclear Information System (INIS)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J.

    2012-01-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000 R nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is

  1. AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel

  2. MDEP AP1000WG Design-Specific Common Position CP-AP1000WG-02. Common position addressing Fukushima Daiichi NPP accident-related issues

    International Nuclear Information System (INIS)

    2016-09-01

    A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami had not been taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive release. Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of nuclear power plants in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of AP1000 R plants are engaged in similar activities. The MDEP AP1000 R Working Group (AP1000 WG) members consist of members from Canada, China, the United Kingdom and the United States. Since the regulatory review of their AP1000 R applications have not been completed by all of these Countries yet, this paper identifies common preliminary approaches to address potential safety improvements for AP1000 R plants as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. In seeking common position, regulators will provide input to this paper to reflect their safety conclusions regarding the AP1000 R design and how the design could be enhanced to address Fukushima Daiichi issues. The common preliminary approaches are organized into five sections

  3. Original earthquake design basis in light of recent seismic hazard studies

    International Nuclear Information System (INIS)

    Petrovski, D.

    1993-01-01

    For the purpose of conceiving the framework within which efforts have been made in the eastern countries to construct earthquake resistant nuclear power plants, a review of the development and application of the seismic zoning map of USSR is given. The normative values of seismic intensity and acceleration are discussed from the aspect of recent probabilistic seismic hazard studies. To that effect, presented briefly in this paper is the methodology of probabilistic seismic hazard analysis. (author)

  4. SISPRO: research and development on the seismic effects attenuation with depth for the seismic design of a long term nuclear waste disposal in the subsurface domain

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, D.; Bossu, R.; Le Piver, F.; Desveaux, F.; Seys, C.; Bouchez, J

    2001-07-01

    In the framework of the 1991/12/30 french law on the management of the nuclear industry waste, the French Atomic Energy Commission (C.E.A.) studies potential benefits against seismic risk of the subsurface domain for the design of an interim storage installation. Indeed, few damage has been observed on subsurface structures during large earthquakes which implied major destructive effects on surface buildings, as during the 1995 Kobe earthquake. However, knowledge on seismic design for subsurface facilities is mainly based on empirical know- how, without satisfactory scientific background which could allow characterization of any given site seismic wave attenuation with depth. The SISPRO program intends to fulfill this lack with two complementary research axis: data acquisition and analysis at several depths and in/on mountain topographies on one hand, accurate numerical modeling on the other hand. The latter will be useful for the establishment of a methodology able to predict seismic waves amplitude, depending on the geotechnical site characteristics and depth. Data analysis which has already been made, such as attenuation laws with several sites data and depth as a parameter, will be depicted. Numerical modeling is based on a 3-D finite differences method able to carry computation of synthetics in any kind of geology. A specific research program is devoted to the case when a topography is present. Numerical results show an attenuation which is smaller than the observed one. This implies that the introduction of a strong gradient in the surface layers properties is probably necessary. Perspectives of the SISPRO program until 2006 will be presented, such as strong motion modeling and how to take into account soil-structure interaction. (author)

  5. SISPRO: research and development on the seismic effects attenuation with depth for the seismic design of a long term nuclear waste disposal in the subsurface domain

    International Nuclear Information System (INIS)

    Rodriguez, D.; Bossu, R.; Le Piver, F.; Desveaux, F.; Seys, C.; Bouchez, J.

    2001-01-01

    In the framework of the 1991/12/30 french law on the management of the nuclear industry waste, the French Atomic Energy Commission (C.E.A.) studies potential benefits against seismic risk of the subsurface domain for the design of an interim storage installation. Indeed, few damage has been observed on subsurface structures during large earthquakes which implied major destructive effects on surface buildings, as during the 1995 Kobe earthquake. However, knowledge on seismic design for subsurface facilities is mainly based on empirical know- how, without satisfactory scientific background which could allow characterization of any given site seismic wave attenuation with depth. The SISPRO program intends to fulfill this lack with two complementary research axis: data acquisition and analysis at several depths and in/on mountain topographies on one hand, accurate numerical modeling on the other hand. The latter will be useful for the establishment of a methodology able to predict seismic waves amplitude, depending on the geotechnical site characteristics and depth. Data analysis which has already been made, such as attenuation laws with several sites data and depth as a parameter, will be depicted. Numerical modeling is based on a 3-D finite differences method able to carry computation of synthetics in any kind of geology. A specific research program is devoted to the case when a topography is present. Numerical results show an attenuation which is smaller than the observed one. This implies that the introduction of a strong gradient in the surface layers properties is probably necessary. Perspectives of the SISPRO program until 2006 will be presented, such as strong motion modeling and how to take into account soil-structure interaction. (author)

  6. Analysis of design floor response spectra and testing of the electrical systems

    International Nuclear Information System (INIS)

    Ambriashvili, Y.

    1996-01-01

    This report covers the following activities as foreseen according to the working plan of 'Atmoenergoproject': analysis of calculated floor response spectra used during the design of Kozloduy NPP and comparison with other spectra recommended for this NPP; analysis of floor response spectrum for the most important systems (reactor, main coolant loop, electrical systems); tests of main electrical systems and analysis of the results on seismic stability of those systems. Results of the response spectra analysis are given, some of the electrical systems are identified by the Kozloduy authorities to be analyzed in future according to the results of the test on seismicity

  7. Seismic analysis and design of steel beam - thick slab floor systems

    International Nuclear Information System (INIS)

    Reed, P.W.

    1981-01-01

    This paper presents a method for seismic analysis and design of floor systems composed of thick reinforced concrete slabs supported by steel beams. The response spectrum modal analysis is used to determine the dynamic response of an orthotropic finite element model. An approximate approach to find the fundamental frequency is explained, allowing an actual acceleration to be determined. The fundamental mode is found to be a major portion of the overall response, whereas the secondary modes are shown to result in a very small portion of the overall response. Dynamic multipliers for the fundamental mode and significant secondary modes are given for several typical floor layouts. These would be used to find equivalent static stress resultants which are used to design the floor. (orig.)

  8. Investigation of optimal seismic design methodology for piping systems supported by elasto-plastic dampers. Part 1. Evaluation functions

    International Nuclear Information System (INIS)

    Ito, Tomohiro; Michiue, Masashi; Fujita, Katsuhisa

    2009-01-01

    In this study, the optimal seismic design methodology that can consider the structural integrity of not only the piping systems but also elasto-plastic supporting devices is developed. This methodology employs a genetic algorithm and can search the optimal conditions such as the supporting location, capacity and stiffness of the supporting devices. Here, a lead extrusion damper is treated as a typical elasto-plastic damper. Four types of evaluation functions are considered. It is found that the proposed optimal seismic design methodology is very effective and can be applied to the actual seismic design for piping systems supported by elasto-plastic dampers. The effectiveness of the evaluation functions is also clarified. (author)

  9. NPP service life management

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    Problems of NPP service life management and service life prolongation are reviewed. Methods for the prolongation of the French NPP service life are discussed, priority directions of nuclear block service life management in regard to aging in the context of the European program of investigation into the materials aging are identified. Questions of the provision of the 60 years service life of the Mihama 1 block (Japan) and decision of the problem of the control equipment aging in Great Britain are discussed. Situation with the prolongation of licenses on the NPP operation in the USA and Spain is considered [ru

  10. Temporary and Long Term Design Provisions Taken on the French NPP Fleet to Cope with Extended Station Black out in case of Rare and Severe External Events

    International Nuclear Information System (INIS)

    Dupuy, Patricia; Delafond, Carine; Dubois, Alexandre

    2015-01-01

    Following the events at Fukushima, the Institute for Radiological Protection and Nuclear Safety (IRSN) has been strongly involved in a series of reviews related to the robustness of French nuclear power plants in case of 'rare and severe' external hazards. These reviews included in particular the 'stress tests' performed in 2011 as required by the European Commission. Those reviews, and the proposal made by EDF to reinforce NPPs robustness in such situation, led to the introduction of the concept of a hardened safety core (HSC) to avoid massive releases and prolonged effects in the environment in case of rare and severe natural hazards. This concept will be explained in the paper and the new specific electrical equipment as well as the interfaces with the existing electrical distribution required to implement this HSC will be explained. As the detailed design, manufacturing and installation of the HSC in all NPP sites will take several years, temporary measures have been adopted. This paper will also present the electrical sources and the distribution related to those temporary measures. The specific situation of the new built EPR reactor in Flamanville is also addressed. Lastly, in complement to the above on-site design provisions, a Nuclear Rapid Response Force has been set up by EDF to bring off-site support to French NPPs in case of emergency. The paper will describe the type of electrical equipment to be delivered and the principle for distributing the electrical power to the required loads. (authors)

  11. Spatial correlation analysis of seismic noise for STAR X-ray infrastructure design

    Science.gov (United States)

    D'Alessandro, Antonino; Agostino, Raffaele; Festa, Lorenzo; Gervasi, Anna; Guerra, Ignazio; Palmer, Dennis T.; Serafini, Luca

    2014-05-01

    The Italian PON MaTeRiA project is focused on the creation of a research infrastructure open to users based on an innovative and evolutionary X-ray source. This source, named STAR (Southern Europe TBS for Applied Research), exploits the Thomson backscattering process of a laser radiation by fast-electron beams (Thomson Back Scattering - TBS). Its main performances are: X-ray photon flux 109-1010 ph/s, Angular divergence variable between 2 and 10 mrad, X-ray energy continuously variable between 8 keV and 150 keV, Bandwidth ΔE/E variable between 1 and 10%, ps time resolved structure. In order to achieve this performances, bunches of electrons produced by a photo-injector are accelerated to relativistic velocities by a linear accelerator section. The electron beam, few hundreds of micrometer wide, is driven by magnetic fields to the interaction point along a 15 m transport line where it is focused in a 10 micrometer-wide area. In the same area, the laser beam is focused after being transported along a 12 m structure. Ground vibrations could greatly affect the collision probability and thus the emittance by deviating the paths of the beams during their travel in the STAR source. Therefore, the study program to measure ground vibrations in the STAR site can be used for site characterization in relation to accelerator design. The environmental and facility noise may affect the X-ray operation especially if the predominant wavelengths in the microtremor wavefield are much smaller than the size of the linear accelerator. For wavelength much greater, all the accelerator parts move in phase, and therefore also large displacements cannot generate any significant effect. On the other hand, for wavelengths equal or less than half the accelerator size several parts could move in phase opposition and therefore small displacements could affect its proper functioning. Thereafter, it is important to characterize the microtremor wavefield in both frequencies and wavelengths domains

  12. Leningrad NPP and energetics of north-western Russia

    International Nuclear Information System (INIS)

    Belov, I.

    2000-01-01

    Problems of Leningrad NPP operating units modernization, their design service life finishing by year of 2010, are discussed. To assure safe operation of unit 1 investments in the amount of 30 mln. dol are necessary. Estimations suggest economic efficiency of the measures, permitting saving of 300 mln. dol worth of gas. Unfortunately, without a rise in tariff for electric power produced by NPP it seems impossible. It is recommended that substantiated tariffs are set for electric power produced by NPP starting from January, 2000. The measure is indispensable for raising investment funds intended for operating NPP modernization [ru

  13. On the Need for Reliable Seismic Input Assessment for Optimized Design and Retrofit of Seismically Isolated Civil and Industrial Structures, Equipment, and Cultural Heritage

    Science.gov (United States)

    Martelli, Alessandro

    2011-01-01

    Based on the experience of recent violent earthquakes, the limits of the methods that are currently used for the definition of seismic hazard are becoming more and more evident to several seismic engineers. Considerable improvement is felt necessary not only for the seismic classification of the territory (for which the probabilistic seismic hazard assessment—PSHA—is generally adopted at present), but also for the evaluation of local amplification. With regard to the first item, among others, a better knowledge of fault extension and near-fault effects is judged essential. The aforesaid improvements are particularly important for the design of seismically isolated structures, which relies on displacement. Thus, such a design requires an accurate definition of the maximum value of displacement corresponding to the isolation period, and a reliable evaluation of the earthquake energy content at the low frequencies that are typical of the isolated structures, for the site and ground of interest. These evaluations shall include possible near-fault effects even in the vertical direction; for the construction of high-risk plants and components and retrofit of some cultural heritage, they shall be performed for earthquakes characterized by very long return periods. The design displacement shall not be underestimated, but neither be excessively overestimated, at least when using rubber bearings in the seismic isolation (SI) system. In fact, by decreasing transverse deformation of such SI systems below a certain value, their horizontal stiffness increases. Thus, should a structure (e.g. a civil defence centre, a masterpiece, etc.) protected in the aforesaid way be designed to withstand an unnecessarily too large earthquake, the behaviour of its SI system will be inadequate (i.e. it will be too stiff) during much more frequent events, which may really strike the structure during its life. Furthermore, since SI can be used only when the room available to the structure

  14. Seismic analysis during development stage of CANDU Model 2 fueling machine design

    International Nuclear Information System (INIS)

    Lee, L.S.S.; Mansfield, R.A.

    1989-01-01

    The CANDU Model 3 is a new small reactor presently being designed. This reactor is 450 MWe, and as with current operating CANDU's, is based on a heavy water moderated and cooled system using on-power fuelling for the once-through natural uranium fuel cycle. The CANDU 3 Standard plant is designed to be adaptable to a range of world-wide site conditions, i.e. for a peak ground acceleration of 0.3 g and a wide range of soft, medium and hard foundation medium properties. Consequently, a conservatism in the design of structure and equipment is accounted by using enveloped floor response spectra generated by the soil-structure interaction analysis. Seismic qualification of the fuelling machine (F/M) and its support structure are an essential design requirement for maintaining the integrity of the reactor coolant heat transport system (HTS) pressure boundary and the service ports penetrating the containment structure during on-power fueling. This paper deals with the initial conceptual phase of design where the details of the design are in fundamental outline form only and basic mass distribution plus layout geometry is defined

  15. Recent Seismicity in Texas and Research Design and Progress of the TexNet-CISR Collaboration

    Science.gov (United States)

    Hennings, P.; Savvaidis, A.; Rathje, E.; Olson, J. E.; DeShon, H. R.; Datta-Gupta, A.; Eichhubl, P.; Nicot, J. P.; Kahlor, L. A.

    2017-12-01

    The recent increase in the rate of seismicity in Texas has prompted the establishment of an interdisciplinary, interinstitutional collaboration led by the Texas Bureau of Economic Geology which includes the TexNet Seismic Monitoring and Research project as funded by The State of Texas (roughly 2/3rds of our funding) and the industry-funded Center for Integrated Seismicity Research (CISR) (1/3 of funding). TexNet is monitoring and cataloging seismicity across Texas using a new backbone seismic network, investigating site-specific earthquake sequences by deploying temporary seismic monitoring stations, and conducting reservoir modeling studies. CISR expands TexNet research into the interdisciplinary realm to more thoroughly study the factors that contribute to seismicity, characterize the associated hazard and risk, develop strategies for mitigation and management, and develop methods of effective communication for all stakeholders. The TexNet-CISR research portfolio has 6 themes: seismicity monitoring, seismology, geologic and hydrologic description, geomechanics and reservoir modeling, seismic hazard and risk assessment, and seismic risk social science. Twenty+ specific research projects span and connect these themes. We will provide a synopsis of research progress including recent seismicity trends in Texas; Fort Worth Basin integrated studies including geological modeling and fault characterization, fluid injection data syntheses, and reservoir and geomechanical modeling; regional ground shaking characterization and mapping, infrastructure vulnerability assessment; and social science topics of public perception and information seeking behavior.

  16. Engineering safety review mission Krsko NPP external events PSA. Ljubljana, Slovenia 19-23 February 1996. Final report

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Smith, P.

    1996-01-01

    Within the scope of the TC Project RER/9/035, a review mission visited Ljubljana, Slovenia, 19-23 February 1996. Two outside experts, Messrs. R.J. Budnitz (USA) and Paul Smith (USA), as well as a staff member, A. Guerpinar (ESS-NSNI) took part in the review. The purpose of the mission was to assist the Slovenian Nuclear Safety Administration to review the external events PSA prepared by Krsko NPP consultants Westinghouse Energy Systems Europe and EQE International. Another seismic safety review was performed concurrently in Ljubljana involving the investigations in relation to the tectonic stability and reassessment of the design basis ground motion characterization for the Krsko NPP site

  17. Akkuyu NPP – the first Turkish NPP. The new history of the project

    International Nuclear Information System (INIS)

    Tzocheva, V.

    2012-01-01

    An overview is given to the Turkish energy sector and nuclear power plans. The project for the construction of the first NPP in Turkey is presented. The general parameters of the Project are: CAPEX: $ 20 bln; Project design: NPP-2006; (VVER- 1200); Number of units: 4; Total capacity: 4 800 MW; Construction period: 2014 – 2023; PPA period; 15 years, fixed price terms. An account of the activities during 2011, the Worley Parsons participation are presented and a tentative project schedule is given

  18. Review of Seismic Evaluation Methodologies for Nuclear Power Plants Based on a Benchmark Exercise

    International Nuclear Information System (INIS)

    2013-11-01

    Niigataken-chuetsu-oki (NCO) earthquake (Mw = 6.6) occurred on 16 July 2007 and affected the Kashiwazaki-Kariwa (K-K) NPP in Japan. Although there was significant loss of main shock data due to transmission problems, a significant number of instruments were still able to measure the acceleration at different locations in soil (boreholes) and in structures at the K-K NPP during the main shock and the aftershocks. The availability of all these instrumental data provided an excellent background for initiating a benchmarking exercise known as the KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment (KARISMA). The main objective of the KARISMA benchmark exercise is to study a comparison between analytical seismic response versus real response of selected structure, system and components (SSCs) of K-K NPP Unit 7. The KARISMA benchmark exercise includes benchmarking the analytical tools and numerical simulation techniques used for predicting seismic response of NPP structures (in linear and non-linear ranges), site response, soil-structure interaction phenomena, seismic response of piping systems, 'sloshing' in the spent fuel pool and buckling of tanks. The benchmark is primarily based on data provided by Tokyo Electric Power Company (TEPCO). It is not linked to the seismic re-evaluation of K-K NPP carried out by TEPCO. Twenty-one organizations, comprising researchers, operating organizations, regulatory authorities, vendors and technical support organizations from 14 countries, participated in the benchmarking exercises. This publication, including a CD-ROM, summarizes the analyses of the main results of the benchmarking exercise for the K-K NPP reactor building (including static and modal analyses of the fixed base model, soil column analyses, analyses of the soil-structure models and margin assessment of the K-K NPP reactor building), the analyses of the main results of the benchmarking exercise for the residual heat removal piping system (including

  19. Displacement-Based Seismic Design Procedure for Framed Buildings with Dissipative Braces Part II: Numerical Results

    International Nuclear Information System (INIS)

    Mazza, Fabio; Vulcano, Alfonso

    2008-01-01

    For a widespread application of dissipative braces to protect framed buildings against seismic loads, practical and reliable design procedures are needed. In this paper a design procedure based on the Direct Displacement-Based Design approach is adopted, assuming the elastic lateral storey-stiffness of the damped braces proportional to that of the unbraced frame. To check the effectiveness of the design procedure, presented in an associate paper, a six-storey reinforced concrete plane frame, representative of a medium-rise symmetric framed building, is considered as primary test structure; this structure, designed in a medium-risk region, is supposed to be retrofitted as in a high-risk region, by insertion of diagonal braces equipped with hysteretic dampers. A numerical investigation is carried out to study the nonlinear static and dynamic responses of the primary and the damped braced test structures, using step-by-step procedures described in the associate paper mentioned above; the behaviour of frame members and hysteretic dampers is idealized by bilinear models. Real and artificial accelerograms, matching EC8 response spectrum for a medium soil class, are considered for dynamic analyses

  20. Specific issues and proposals in aseismic design technologies (seismic isolation technologies)

    International Nuclear Information System (INIS)

    Fujita, Satoshi

    2000-01-01

    It is examined among engineers to control vibration of buildings and constructions formed by earthquake, and at present various vibration control techniques are in actual use. A vibration isolating structure passing through earthquake, and vibration controlling due to wind are its typical ones, which have been recently and rapidly supplied to actual use through a chance that laminated rubber was researched and developed for a vibration isolation supporting materials capable of supplying to actual use about 15 years ago. However, the active addition mass type vibration controller is not adequate to large earthquake countermeasure from points of addition mass size, drive variation, and limit of control power. For a vibration controller suitable for this aim an energy absorber (damper) of a type set between layers of constructions at present is the most predominant, of which various types are earnestly under research and development. Here were explained on earthquake and its energy, seismic resistant design, vibration isolation structure, and so forth. (G.K.)