Seismic evaluation of piping systems using screening criteria

International Nuclear Information System (INIS)

Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

1994-01-01

This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

A procedure for seismic risk reduction in Campania Region

International Nuclear Information System (INIS)

Zuccaro, G.; Palmieri, M.; Cicalese, S.; Grassi, V.; Rauci, M.; Maggio, F.

2008-01-01

The Campania Region has set and performed a peculiar procedure in the field of seismic risk reduction. Great attention has been paid to public strategic buildings such as town halls, civil protection buildings and schools. The Ordinance 3274 promulgate in the 2004 by the Italian central authority obliged the owners of strategic buildings to perform seismic analyses within 2008 in order to check the safety of the structures and the adequacy to the use. In the procedure the Campania region, instead of the local authorities, ensure the complete drafting of seismic checks through financial resources of the Italian Government. A regional scientific technical committee has been constituted, composed of scientific experts, academics in seismic engineering. The committee has drawn up guidelines for the processing of seismic analyses. At the same time, the Region has issued a public competition to select technical seismic engineering experts to appoint seismic analysis in accordance with guidelines. The scientific committee has the option of requiring additional documents and studies in order to approve the safety checks elaborated. The Committee is supported by a technical and administrative secretariat composed of a group of expert in seismic engineering. At the moment several seismic safety checks have been completed. The results will be presented in this paper. Moreover, the policy to mitigate the seismic risk, set by Campania region, was to spend the most of the financial resources available on structural strengthening of public strategic buildings rather than in safety checks. A first set of buildings of which the response under seismic action was already known by data and studies of vulnerability previously realised, were selected for immediate retrofitting designs. Secondly, an other set of buildings were identified for structural strengthening. These were selected by using the criteria specified in the Guide Line prepared by the Scientific Committee and based on

A procedure for seismic risk reduction in Campania Region

Science.gov (United States)

Zuccaro, G.; Palmieri, M.; Maggiò, F.; Cicalese, S.; Grassi, V.; Rauci, M.

2008-07-01

The Campania Region has set and performed a peculiar procedure in the field of seismic risk reduction. Great attention has been paid to public strategic buildings such as town halls, civil protection buildings and schools. The Ordinance 3274 promulgate in the 2004 by the Italian central authority obliged the owners of strategic buildings to perform seismic analyses within 2008 in order to check the safety of the structures and the adequacy to the use. In the procedure the Campania region, instead of the local authorities, ensure the complete drafting of seismic checks through financial resources of the Italian Government. A regional scientific technical committee has been constituted, composed of scientific experts, academics in seismic engineering. The committee has drawn up guidelines for the processing of seismic analyses. At the same time, the Region has issued a public competition to select technical seismic engineering experts to appoint seismic analysis in accordance with guidelines. The scientific committee has the option of requiring additional documents and studies in order to approve the safety checks elaborated. The Committee is supported by a technical and administrative secretariat composed of a group of expert in seismic engineering. At the moment several seismic safety checks have been completed. The results will be presented in this paper. Moreover, the policy to mitigate the seismic risk, set by Campania region, was to spend the most of the financial resources available on structural strengthening of public strategic buildings rather than in safety checks. A first set of buildings of which the response under seismic action was already known by data and studies of vulnerability previously realised, were selected for immediate retrofitting designs. Secondly, an other set of buildings were identified for structural strengthening. These were selected by using the criteria specified in the Guide Line prepared by the Scientific Committee and based on

Rethinking ASME III seismic analysis for piping operability evaluations

International Nuclear Information System (INIS)

Adams, T.M.; Stevenson, J.D.

1994-01-01

It has been recognized since the mid 1980's that there are very large seismic margins to failure for nuclear piping systems when designed using current industry practice, design criteria, and methods. As a result of this realization there are or have been approximately eighteen initiatives within the ASME , Boiler and Pressure Vessel Code Section III, Division 1, in the form of proposed code cases and proposed code text changes designed to reduce these failure margins to more realistic values. For the most part these initiatives have concentrated on reclassifying seismic inertia stresses in the piping as secondary and increasing the allowable stress limits permitted by Section III of the ASME, Boiler Code. This paper focuses on the application of non-linear spectral analysis methods as a method to reduce the input seismic demand determination and thereby reduce the seismic failure margins. The approach is evaluated using the ASME Boiler Pressure Vessel Code Section III Subgroup on Design benchmark procedure as proposed by the Subgroup's Special Task Group on Integrated Piping Criteria. Using this procedure, criteria are compared to current code criterion and analysis methods, and several other of the currently proposed Boiler and Pressure Vessel, Section III, changes. Finally, the applicability of the non-linear spectral analysis to continued Safe Operation Evaluations is reviewed and discussed

U.S. experience in seismic re-evaluation and verification programs

International Nuclear Information System (INIS)

Stevenson, J.D.

1995-01-01

The purpose of this paper is to present a summary of the development of a seismic re-evaluation program for older nuclear power plants in the U.S. The principal focus of this reevaluation is the use of actual strong motion earthquake response data for structures and mechanical and electrical systems and components. These data are supplemented by generic shake table test results. Use of this type of seismic re-evaluation has led to major cost reductions as compared to more conventional analytical and component specific testing procedures. (author)

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)

Seismic evaluation of buildings in the Eastern and Central United States

International Nuclear Information System (INIS)

Malley, J.O.; Poland, C.D.

1991-01-01

The vast majority of the existing buildings in the Central and Eastern United States have not been designed to resist seismic forces, even though it is becoming widely accepted that there is a potential for damaging earthquakes in these regions for the country. These buildings, therefore, may constitute a serious threat to life safety in the event of a major earthquake. The ATC-14 procedure for the seismic evaluation of existing buildings has begun to gain wide acceptance since its publication in 1987. The National Center for Earthquake Engineering Research (NCEER) funded a project to critically assess the applicability of ATC-14 to buildings in the Eastern and Central United States. This NCEER project developed a large volume of recommended modifications to ATC-14 which are intended to improve the modifications to ATC-14 procedure's recommendations for the seismic evaluation of buildings in regions of low siesmicity. NCEER is sponsoring a second project which will produce a separate document for the seismic evaluation of existing buildings which specifically focuses on structures in these areas of the country. This report, which should be completed in 1991, will provide a valuable tool for practicing engineers performing these evaluations in the Eastern and Central United States. This paper will present the results of these NCEER projects and introduce the revised ATC-14 methodology

Seismic evaluation of reinforced masonry walls

International Nuclear Information System (INIS)

Kelly, T.E.; Button, M.R.; Mayes, R.L.

1984-01-01

Masonry walls in operating nuclear plants are in many cases found to be overstressed in terms of allowable stresses when evaluated using current seismic design criteria. However, experimental evidence exists indicating that reinforced masonry walls have a considerable margin between the load levels at which allowable stresses are exceeded and the load levels at which structural distress and loss of function occurs. This paper presents a methodology which allows the actual capacity of reinforced masonry walls under seismic loading to be quantified. The methodology is based on the use of non-linear dynamic analyses and incorporates observed hysteretic behavior for both in-plane and out-of-plane response. Experimental data is used to develop response parameters and to validate the results predicted by the models. Criteria have been concurrently developed to evaluate the deformations and material performance in the walls to ensure adequate margins of safety for the required function. An example of the application of these procedures is provided

Seismic qualification of equipment

International Nuclear Information System (INIS)

Heidebrecht, A.C.; Tso, W.K.

1983-03-01

This report describes the results of an investigation into the seismic qualification of equipment located in CANDU nuclear power plants. It is particularly concerned with the evaluation of current seismic qualification requirements, the development of a suitable methodology for the seismic qualification of safety systems, and the evaluation of seismic qualification analysis and testing procedures

Methodology to evaluate the site standard seismic motion to a nuclear facility

International Nuclear Information System (INIS)

Soares, W.A.

1983-01-01

For the seismic design of nuclear facilities, the input motion is normally defined by the predicted maximum ground horizontal acceleration and the free field ground response spectrum. This spectrum is computed on the basis of records of strong motion earthquakes. The pair maximum acceleration-response spectrum is called the site standard seismic motion. An overall view of the subjects involved in the determination of the site standard seismic motion to a nuclear facility is presented. The main topics discussed are: basic principles of seismic instrumentation; dynamic and spectral concepts; design earthquakes definitions; fundamentals of seismology; empirical curves developed from prior seismic data; available methodologies and recommended procedures to evaluate the site standard seismic motion. (Author) [pt

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

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 ...

Seismic isolation - efficient procedure for seismic response assessement

International Nuclear Information System (INIS)

Zamfir, M. A.; Androne, M.

2016-01-01

The aim of this analysis is to reduce the dynamic response of a structure. The seismic isolation solution must take into consideration the specific site ground motion. In this paper will be presented results obtained by applying the seismic isolation method. Based on the obtained results, important conclusions can be outlined: the seismic isolation device has the ability to reduce seismic acceleration of the seismic isolated structure to values that no longer present a danger to people and environment; the seismic isolation solution is limiting devices deformations to safety values for ensuring structural integrity and stability of the entire system; the effective seismic energy dissipation and with no side effects both for the seismic isolated building and for the devices used, and the return to the initial position before earthquake occurence are obtained with acceptable permanent displacement. (authors)

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.)

Evaluation of seismic hazards for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

The main objective of this Safety Guide is to provide recommendations on how to determine the ground motion hazards for a plant at a particular site and the potential for surface faulting, which could affect the feasibility of construction and safe operation of a plant at that site. The guidelines and procedures presented in this Safety Guide can appropriately be used in evaluations of site suitability and seismic hazards for nuclear power plants in any seismotectonic environment. The probabilistic seismic hazard analysis recommended in this Safety Guide also addresses the needs for seismic hazard analysis of external event PSAs conducted for nuclear power plants. Many of the methods and processes described may also be applicable to nuclear facilities other than power plants. Other phenomena of permanent ground displacement (liquefaction, slope instability, subsidence and collapse) as well as the topic of seismically induced flooding are treated in Safety Guides relating to foundation safety and coastal flooding. Recommendations of a general nature are given in Section 2. Section 3 discusses the acquisition of a database containing the information needed to evaluate and address all hazards associated with earthquakes. Section 4 covers the use of this database for construction of a seismotectonic model. Sections 5 and 6 review ground motion hazards and evaluations of the potential for surface faulting, respectively. Section 7 addresses quality assurance in the evaluation of seismic hazards for nuclear power plants

Comparison of seismic margin assessment and probabilistic risk assessment in seismic IPE

International Nuclear Information System (INIS)

Reed, J.W.; Kassawara, R.P.

1993-01-01

A comparison of technical requirements and managerial issues between seismic margin assessment (SMA) and seismic probabilistic risk assessment (SPRA) in a seismic Individual Plant Examination (IPE) is presented and related to requirements for an Unresolved Safety Issue (USI) A-46 review which is required for older nuclear power plants. Advantages and disadvantages are discussed for each approach. Technical requirements reviewed for a seismic IPE include: scope of plants covered, seismic input, scope of review, selection of equipment, required experience and training of engineers, walkdown procedure, evaluation of components, relay review, containment review, quality assurance, products, documentation requirements, and closure procedure. Managerial issues discussed include regulatory acceptability, compatibility with seismic IPE, compliance with seismic IPE requirements, ease of use by utilities, and relative cost

Burar seismic station: evaluation of seismic performance

International Nuclear Information System (INIS)

Ghica, Daniela; Popa, Mihaela

2005-01-01

A new seismic monitoring system, the Bucovina Seismic Array (BURAR), has been established since July 2002, in the Northern part of Romania, in a joint effort of the Air Force Technical Applications Center, USA, and the National Institute for Earth Physics (NIEP), Romania. The small-aperture array consists of 10 seismic sensors (9 vertical short-period and one three-component broad band) located in boreholes and distributed in a 5 x 5 km 2 area. At present, the seismic data are continuously recorded by the BURAR and transmitted in real-time to the Romanian National Data Center in Bucharest and National Data Center of the USA, in Florida. Based on the BURAR seismic information gathered at the National Data Center, NIEP (ROM N DC), in the August 2002 - December 2004 time interval, analysis and statistical assessments were performed. Following the preliminary processing of the data, several observations on the global performance of the BURAR system were emphasized. Data investigation showed an excellent efficiency of the BURAR system particularly in detecting teleseismic and regional events. Also, a statistical analysis for the BURAR detection capability of the local Vrancea events was performed in terms of depth and magnitude for the year 2004. The high signal detection capability of the BURAR resulted, generally, in improving the location solutions for the Vrancea seismic events. The location solution accuracy is enhanced when adding BURAR recordings, especially in the case of low magnitude events (recorded by few stations). The location accuracy is increased, both in terms of constraining hypocenter depth and epicentral coordinates. Our analysis certifies the importance of the BURAR system in NIEP efforts to elaborate seismic bulletins. Furthermore, the specific procedures for array data processing (beam forming, f-k analysis) increase significantly the signal-to-noise ratio by summing up the coherent signals from the array components, and ensure a better accuracy

Integrated system for seismic evaluations

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulos, A.J.; Miller, C.A.; Costantino, C.J.; Graves, H.

1989-01-01

This paper describes the various features of the seismic module of the CARES system (computer analysis for rapid evaluation of structures). This system was developed to perform rapid evaluations of structural behavior and capability of nuclear power plant facilities. The CARES is structural in a modular format. Each module performs a specific type of analysis i.e., static or dynamic, linear or nonlinear, etc. This paper describes the features of the seismic module in particular. The development of the seismic modules of the CARES system is based on an approach which incorporates major aspects of seismic analysis currently employed by the industry into an integrated system that allows for carrying out interactively computations of structural response to seismic motions. The code operates on a PC computer system and has multi-graphics capabilities

Seismic evaluation of existing nuclear facilities. Proceedings

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

Programmes for re-evaluation and upgrading of safety of existing nuclear facilities are presently under way in a number of countries around the world. An important component of these programmes is the re-evaluation of the seismic safety through definition of new seismic parameters at the site and evaluation of seismic capacity of structures, equipment and distribution systems following updated information and criteria. The Seminar is intended to provide a forum for the exchange of information and discussion of the state-of-the-art on seismic safety of nuclear facilities in operation or under construction. Both analytical and experimental techniques for the evaluation of seismic capacity of structures, equipment and distribution systems are discussed. Full scale and field tests of structures and components using shaking tables, mechanical exciters, explosive and shock tests, and ambient vibrations are included in the seminar programme with emphasis on recent case histories. Presentations at the Seminar also include analytical techniques for the determination of dynamic properties of soil-structure systems from experiments as well as calibration of numerical models. Methods and criteria for seismic margin assessment based on experience data obtained from the behaviour of structures and components in real earthquakes are discussed. Guidelines for defining technical requirements for capacity re-evaluation (i.e. acceptable behaviour limits and design and implementation of structure and components upgrades are also presented and discussed. The following topics were covered during 7 sessions: earthquake experience and seismic re-evaluation; country experience in seismic re-evaluation programme; generic WWER studies; analytical methods for seismic capacity re-evaluation; experimental methods for seismic capacity re-evaluation; case studies.

Seismic evaluation of existing nuclear facilities. Proceedings

International Nuclear Information System (INIS)

1995-01-01

Programmes for re-evaluation and upgrading of safety of existing nuclear facilities are presently under way in a number of countries around the world. An important component of these programmes is the re-evaluation of the seismic safety through definition of new seismic parameters at the site and evaluation of seismic capacity of structures, equipment and distribution systems following updated information and criteria. The Seminar is intended to provide a forum for the exchange of information and discussion of the state-of-the-art on seismic safety of nuclear facilities in operation or under construction. Both analytical and experimental techniques for the evaluation of seismic capacity of structures, equipment and distribution systems are discussed. Full scale and field tests of structures and components using shaking tables, mechanical exciters, explosive and shock tests, and ambient vibrations are included in the seminar programme with emphasis on recent case histories. Presentations at the Seminar also include analytical techniques for the determination of dynamic properties of soil-structure systems from experiments as well as calibration of numerical models. Methods and criteria for seismic margin assessment based on experience data obtained from the behaviour of structures and components in real earthquakes are discussed. Guidelines for defining technical requirements for capacity re-evaluation (i.e. acceptable behaviour limits and design and implementation of structure and components upgrades are also presented and discussed. The following topics were covered during 7 sessions: earthquake experience and seismic re-evaluation; country experience in seismic re-evaluation programme; generic WWER studies; analytical methods for seismic capacity re-evaluation; experimental methods for seismic capacity re-evaluation; case studies

Status report on seismic re-evaluation

International Nuclear Information System (INIS)

1998-01-01

In May 1997, a meeting of the PWG 3 Sub Group on the Seismic Behaviour of Structures agreed several priority objectives, of which one was the production of a status report on seismic re-evaluation. Seismic re-evaluation is identified as the process of carrying out a re-assessment of the safety of existing nuclear power plants for a specified seismic hazard. This may be necessary when no seismic hazard was considered in the original design of the plant, the relevant codes and regulations have been revised, the seismic hazard for the site has been re-assessed or there is a need to assess the capacity of the plant for severe accident conditions and behaviour beyond the design basis. Re-evaluation may also be necessary to resolve an issue, or to assess the impact of new findings or knowledge. A questionnaire on the subject was issued to all members of the Seismic Sub Group in the summer of 1997, and responses to the questionnaire had been received from most members by the end of 1997. This report is based on the responses to the questionnaire, together with comment and discussion within the group. The questionnaire covered the following main topics of interest in relation to seismic re-evaluation: General and Legislative Framework, Overall Approach, Input Definition and Analysis Methods, Scope of Plant and Assessment of As-built Situation, Assessment criteria, Outcome of Re-evaluations, Research. The responses to the questionnaire have been collated and reviewed with the objective of comparing current practice in the field of seismic re-evaluation in member countries, and a number of important points have been identified in relation to the position of seismic re-evaluation in the nuclear power industry throughout the world. It is evident that seismic re-evaluation is a relatively mature process that has been developing for some time, with most countries adopting similar practices, often based on principles which have been developed in the US nuclear industry. Seismic

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

Seismic Hazards in Site Evaluation for Nuclear Installations. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-08-15

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear installations. It supplements the Safety Requirements publication on Site Evaluation for Nuclear Installations. The present publication provides guidance and recommends procedures for the evaluation of seismic hazards for nuclear power plants and other nuclear installations. It supersedes Evaluation of Seismic Hazards for Nuclear Power Plants, IAEA Safety Standards Series No. NS-G-3.3 (2002). In this publication, the following was taken into account: the need for seismic hazard curves and ground motion spectra for the probabilistic safety assessment of external events for new and existing nuclear installations; feedback of information from IAEA reviews of seismic safety studies for nuclear installations performed over the previous decade; collective knowledge gained from recent significant earthquakes; and new approaches in methods of analysis, particularly in the areas of probabilistic seismic hazard analysis and strong motion simulation. In the evaluation of a site for a nuclear installation, engineering solutions will generally be available to mitigate, by means of certain design features, the potential vibratory effects of earthquakes. However, such solutions cannot always be demonstrated to be adequate for mitigating the effects of phenomena of significant permanent ground displacement such as surface faulting, subsidence, ground collapse or fault creep. The objective of this Safety Guide is to provide recommendations and guidance on evaluating seismic hazards at a nuclear installation site and, in particular, on how to determine: (a) the vibratory ground motion hazards, in order to establish the design basis ground motions and other relevant parameters for both new and existing nuclear installations; and (b) the potential for fault displacement and the rate of fault displacement that could affect the feasibility of the site or the safe operation of the installation at

Integrated system for seismic evaluations

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulos, A.J.; Miller, C.A.; Costantino, C.J.; Graves, H.

1989-01-01

This paper describes the various features of the Seismic Module of the CARES system (Computer Analysis for Rapid Evaluation of Structures). This system was developed by Brookhaven National Laboratory (BNL) for the US Nuclear Regulatory Commission to perform rapid evaluations of structural behavior and capability of nuclear power plant facilities. The CARES is structured in a modular format. Each module performs a specific type of analysis i.e., static or dynamic, linear or nonlinear, etc. This paper describes the features of the Seismic Module in particular. The development of the Seismic Module of the CARES system is based on an approach which incorporates all major aspects of seismic analysis currently employed by the industry into an integrated system that allows for carrying out interactively computations of structural response to seismic motions. The code operates on a PC computer system and has multi-graphics capabilities. It has been designed with user friendly features and it allows for interactive manipulation of various analysis phases during the seismic design process. The capabilities of the seismic module include (a) generation of artificial time histories compatible with given design ground response spectra, (b) development of Power Spectral Density (PSD) functions associated with the seismic input, (c) deconvolution analysis using vertically propagating shear waves through a given soil profile, and (d) development of in-structure response spectra or corresponding PSD's. It should be pointed out that these types of analyses can also be performed individually by using available computer codes such as FLUSH, SAP, etc. The uniqueness of the CARES, however, lies on its ability to perform all required phases of the seismic analysis in an integrated manner. 5 refs., 6 figs

PROCEDURES FOR SEISMIC CHARACTERIZATION OF TRADITIONAL AND MODERN WOODEN BUILDING TYPES

Directory of Open Access Journals (Sweden)

Ario Ceccotti

2016-03-01

Full Text Available The paper analyzes different wooden buildings types used in past and nowadays to realized low-rise and mid-rise timber structures from the seismic point of view. A preliminary overview about the procedures prescribed by codes for the seismic characterization of the timber building systems is given. Then the definition of the behaviour q-factor in the literature and its relevance in design of structures in seismic areas is treated. Available research methods for estimating the q-factor based on the verification of the nonlinear seismic response of entire buildings by means of experimental tests and numerical simulations are presented and analyzed. The relevance of a proper definition of the yielding limit and of failure condition in the seismic characterization of wooden building systems is treated. Moreover, a comparison between the q-factor estimations obtained using different calculation methods is presented. Lastly, the appropriate q-factor values are given for a reliable and safe seismic design of buildings realized using the examined wooden constructive systems.

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

Science.gov (United States)

Huang, Y.-N.; Whittaker, A.S.; Luco, N.

2011-01-01

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology. ?? 2011 Published by Elsevier B.V.

Some preliminary results of a worldwide seismicity estimation: a case study of seismic hazard evaluation in South America

Directory of Open Access Journals (Sweden)

C. V. Christova

2000-06-01

Full Text Available Global data have been widely used for seismicity and seismic hazard assessment by seismologists. In the present study we evaluate worldwide seismicity in terms of maps of maximum observed magnitude (Mmax, seismic moment (M 0 and seismic moment rate (M 0S. The data set used consists of a complete and homogeneous global catalogue of shallow (h £ 60 km earthquakes of magnitude MS ³ 5.5 for the time period 1894-1992. In order to construct maps of seismicity and seismic hazard the parameters a and b derived from the magnitude-frequency relationship were estimated by both: a the least squares, and b the maximum likelihood, methods. The values of a and b were determined considering circles centered at each grid point 1° (of a mesh 1° ´1° and of varying radius, which starts from 30 km and moves with a step of 10 km. Only a and b values which fulfill some predefined conditions were considered in the further procedure for evaluating the seismic hazard maps. The obtained worldwide M max distribution in general delineates the contours of the plate boundaries. The highest values of M max observed are along the circum-Pacific belt and in the Himalayan area. The subduction plate boundaries are characterized by the largest amount of M 0 , while areas of continental collision are next. The highest values of seismic moment rate (per 1 year and per equal area of 10 000 km 2 are found in the Southern Himalayas. The western coasts of U.S.A., Northwestern Canada and Alaska, the Indian Ocean and the eastern rift of Africa are characterized by high values of M 0 , while most of the Pacific subduction zones have lower values of seismic moment rate. Finally we analyzed the seismic hazard in South America comparing the predicted by the NUVEL1 model convergence slip rate between Nazca and South America plates with the average slip rate due to earthquakes. This consideration allows for distinguishing between zones of high and low coupling along the studied convergence

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

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.

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

Seismic risk evaluation for high voltage air insulated substations

International Nuclear Information System (INIS)

Camensig, Carlo; Bresesti, Luca; Clementel, Stefano; Salvetti, Maurizio

1997-01-01

This paper describes the results of the analytical and experimental activities performed by ISMES for the evaluation of the structural reliability of electrical substations with respect to seismic events. In the following, the reference station is described along with the methods used to define the site seismic input, the analytical and experimental evaluation of the components' fragility curves and the whole station seismic risk evaluation

BUILDING 341 Seismic Evaluation

Energy Technology Data Exchange (ETDEWEB)

Halle, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-06-15

The Seismic Evaluation of Building 341 located at Lawrence Livermore National Laboratory in Livermore, California has been completed. The subject building consists of a main building, Increment 1, and two smaller additions; Increments 2 and 3.

Summary report of seismic PSA of BWR model plant

International Nuclear Information System (INIS)

1999-05-01

This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

International symposium on seismic evaluation of existing nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Orbovic, N.; Bouchon, M. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Vendel, J.; Gelain, T. [IRSN/DPEA/SERAC, 91 - Gif sur Yvette (France)

2003-10-01

revealed a strong influence of the slab plates and a torsion in the structure. Also, masonry in-fills can significantly alter the seismic behavior of the building. The building is being subjected to a very large retrofit program submitted for approval to IRSN and the Nuclear Safety Authority. The acceptance criteria used to define deficiencies and to retrofit the building were design criteria. Evaluation of existing nuclear facility structures must be conducted with the appropriate level of conservatism. Therefore, even though modern performance-based evaluation procedures for conventional structures are available, they must be rigorously examined before they are used to evaluate nuclear facility structures. This paper presents selected results of an evaluation of various nonlinear static and dynamic demand analysis and FEMA-356 capacity measures on a test-bed nuclear facility reinforced concrete frame structure built in 1960's. It is shown that these modern procedures can be used. Furthermore, they revealed deficiencies of the test-bed structure that could not be easily found using the conventional linear-and-elastic evaluation methods. Nevertheless, more work is needed to calibrate these new procedures for the risk reduction levels required for nuclear facilities before they can be used in nuclear facility design and evaluation practice. The test program consisted of mechanical testy performed on reinforced concrete walls subjected to alternating shear loads. The aim of the program was to define crack geometry (length, width, and spacing). Tests were carried out on 3 low-rise reinforced concrete walls with varying percentages of rebars. Each wall was subjected to a sequence of three increasing loads. One of the three walls was loaded to failure. The main results of the program were measurements of displacement as a function of horizontal force (displacements, variations in diagonal lengths), cracking states during the loading cycle, and deformation of rebars. The

A personal computer code for seismic evaluations of nuclear power plant facilities

International Nuclear Information System (INIS)

Xu, J.; Graves, H.

1990-01-01

A wide range of computer programs and modeling approaches are often used to justify the safety of nuclear power plants. It is often difficult to assess the validity and accuracy of the results submitted by various utilities without developing comparable computer solutions. Taken this into consideration, CARES is designed as an integrated computational system which can perform rapid evaluations of structural behavior and examine capability of nuclear power plant facilities, thus CARES may be used by the NRC to determine the validity and accuracy of analysis methodologies employed for structural safety evaluations of nuclear power plants. CARES has been designed to: operate on a PC, have user friendly input/output interface, and have quick turnaround. The CARES program is structured in a modular format. Each module performs a specific type of analysis. The basic modules of the system are associated with capabilities for static, seismic and nonlinear analyses. This paper describes the various features which have been implemented into the Seismic Module of CARES version 1.0. In Section 2 a description of the Seismic Module is provided. The methodologies and computational procedures thus far implemented into the Seismic Module are described in Section 3. Finally, a complete demonstration of the computational capability of CARES in a typical soil-structure interaction analysis is given in Section 4 and conclusions are presented in Section 5. 5 refs., 4 figs

How to eliminate non-damaging earthquakes from the results of a probabilistic seismic hazard analysis (PSHA)-A comprehensive procedure with site-specific application

International Nuclear Information System (INIS)

Kluegel, Jens-Uwe

2009-01-01

The results of probabilistic seismic hazard analyses are frequently presented in terms of uniform hazard spectra or hazard curves with spectral accelerations as the output parameter. The calculation process is based on the evaluation of the probability of exceedance of specified acceleration levels without consideration of the damaging effects of the causative earthquakes. The same applies to the empirical attenuation equations for spectral accelerations used in PSHA models. This makes interpreting and using the results in engineering or risk applications difficult. Uniform hazard spectra and the associated hazard curves may contain a significant amount of contributions of weak, low-energy earthquakes not able to damage the seismically designed structures of nuclear power plants. For the development of realistic engineering designs and for realistic seismic probabilistic risk assessments (seismic PRA) it is necessary to remove the contribution of non-damaging earthquakes from the results of a PSHA. A detailed procedure for the elimination of non-damaging earthquakes based on the CAV (Cumulative Absolute Velocity)-filtering approach was developed and applied to the results of the large-scale PEGASOS probabilistic seismic hazard study for the site of the Goesgen nuclear power plant. The procedure considers the full scope of epistemic uncertainty and aleatory variability present in the PEGASOS study. It involves the development of a set of empirical correlations for CAV and the subsequent development of a composite distribution for the probability of exceedance of the damaging threshold of 0.16 gs. Additionally, a method was developed to measure the difference in the damaging effects of earthquakes of different strengths by the ratio of a power function of ARIAS-intensity or, in the ideal case, by the ratio of the square roots of the associated strong motion durations. The procedure was applied for the update of the Goesgen seismic PRA and for the confirmation of a

Lower bound earthquake magnitude for probabilistic seismic hazard evaluation

International Nuclear Information System (INIS)

McCann, M.W. Jr.; Reed, J.W.

1990-01-01

This paper presents the results of a study that develops an engineering and seismological basis for selecting a lower-bound magnitude (LBM) for use in seismic hazard assessment. As part of a seismic hazard analysis the range of earthquake magnitudes that are included in the assessment of the probability of exceedance of ground motion must be defined. The upper-bound magnitude is established by earth science experts based on their interpretation of the maximum size of earthquakes that can be generated by a seismic source. The lower-bound or smallest earthquake that is considered in the analysis must also be specified. The LBM limits the earthquakes that are considered in assessing the probability that specified ground motion levels are exceeded. In the past there has not been a direct consideration of the appropriate LBM value that should be used in a seismic hazard assessment. This study specifically looks at the selection of a LBM for use in seismic hazard analyses that are input to the evaluation/design of nuclear power plants (NPPs). Topics addressed in the evaluation of a LBM are earthquake experience data at heavy industrial facilities, engineering characteristics of ground motions associated with small-magnitude earthquakes, probabilistic seismic risk assessments (seismic PRAs), and seismic margin evaluations. The results of this study and the recommendations concerning a LBM for use in seismic hazard assessments are discussed. (orig.)

Aspects of the Iea-R1 research reactor seismic evaluation

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1996-01-01

Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

Development and seismic evaluation of the seismic monitoring analysis system for HANARO

International Nuclear Information System (INIS)

Ryu, J. S.; Youn, D. B.; Kim, H. G.; Woo, J. S.

2003-01-01

Since the start of operation, the seismic monitoring system has been utilized for monitoring an earthquake at the HANARO site. The existing seismic monitoring system consists of field sensors and monitoring panel. The analog-type monitoring system with magnetic tape recorder is out-of-date model. In addition, the disadvantage of the existing system is that it does not include signal-analyzing equipment. Therefore, we have improved the analog seismic monitoring system except the field sensors into a new digital Seismic Monitoring Analysis System(SMAS) that can monitor and analyze earthquake signals. To achieve this objective for HANARO, the digital type hardware of the SMAS has been developed. The seismic monitoring and analysis programs that can provide rapid and precise information for an earthquake were developed. After the installation of the SMAS, we carried out the Site Acceptance Test (SAT) to confirm the functional capability of the newly developed system. The results of the SAT satisfy the requirements of the fabrication technical specifications. In addition, the seismic characteristics and structural integrity of the SMAS were evaluated. The results show that the cabinet of SMAS can withstand the effects of seismic loads and remain functional. This new SMAS is operating in the HANARO instrument room to acquire and analyze the signal of an earthquake

Seismic fragility evaluation of unreinforced masonry walls

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.; Reich, M.; Lee, S.K.

1991-01-01

A practical analysis scheme to evaluate the seismic fragility of unreinforced masonry walls which are used at various places in older reactor facilities is presented. Among the several failure modes for such walls, the out-of-plane bending failure is considered to be a major risk contributor in seismic PRA studies. In order to evaluate this failure mode, the use of an equivalent linear approximation method is examined based on comparisons with available test data and nonlinear time history analyses. (author)

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

International symposium on seismic evaluation of existing nuclear facilities

International Nuclear Information System (INIS)

Orbovic, N.; Bouchon, M.; Vendel, J.; Gelain, T.

2003-10-01

strong influence of the slab plates and a torsion in the structure. Also, masonry in-fills can significantly alter the seismic behavior of the building. The building is being subjected to a very large retrofit program submitted for approval to IRSN and the Nuclear Safety Authority. The acceptance criteria used to define deficiencies and to retrofit the building were design criteria. Evaluation of existing nuclear facility structures must be conducted with the appropriate level of conservatism. Therefore, even though modern performance-based evaluation procedures for conventional structures are available, they must be rigorously examined before they are used to evaluate nuclear facility structures. This paper presents selected results of an evaluation of various nonlinear static and dynamic demand analysis and FEMA-356 capacity measures on a test-bed nuclear facility reinforced concrete frame structure built in 1960's. It is shown that these modern procedures can be used. Furthermore, they revealed deficiencies of the test-bed structure that could not be easily found using the conventional linear-and-elastic evaluation methods. Nevertheless, more work is needed to calibrate these new procedures for the risk reduction levels required for nuclear facilities before they can be used in nuclear facility design and evaluation practice. The test program consisted of mechanical testy performed on reinforced concrete walls subjected to alternating shear loads. The aim of the program was to define crack geometry (length, width, and spacing). Tests were carried out on 3 low-rise reinforced concrete walls with varying percentages of rebars. Each wall was subjected to a sequence of three increasing loads. One of the three walls was loaded to failure. The main results of the program were measurements of displacement as a function of horizontal force (displacements, variations in diagonal lengths), cracking states during the loading cycle, and deformation of rebars. The measurements

Seismic fragility evaluation of unreinforced masonry walls

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.; Reich, M.; Lee, S.K.

1991-01-01

A practical analysis scheme to evaluate the seismic fragility of unreinforced masonry walls which are used to various places in older reactor facilities is presented. Among the several failure modes for such walls, the out-of-plane bending failure is considered to be a major risk contributor in seismic PRA studies. In order to evaluate this failure mode, the use of an equivalent linear approximation method is examined based on comparisons with available test data and nonlinear time history analyses. 6 refs., 4 figs., 3 tabs

Aging evaluation of class 1E batteries: Seismic testing

International Nuclear Information System (INIS)

Edson, J.L.

1990-08-01

This report presents the results of a seismic testing program on naturally aged class 1E batteries obtained from a nuclear plant. The testing program is a Phase 2 activity resulting from a Phase 1 aging evaluation of class 1E batteries in safety systems of nuclear power plants, performed previously as a part of the US Nuclear Regulatory Commission's Nuclear Plant Aging Research Program and reported in NUREG/CR-4457. The primary purpose of the program was to evaluate the seismic ruggedness of naturally aged batteries to determine if aged batteries could have adequate electrical capacity, as determined by tests recommended by IEEE Standards, and yet have inadequate seismic ruggedness to provide needed electrical power during and after a safe shutdown earthquake (SSE) event. A secondary purpose of the program was to evaluate selected advanced surveillance methods to determine if they were likely to be more sensitive to the aging degradation that reduces seismic ruggedness. The program used twelve batteries naturally aged to about 14 years of age in a nuclear facility and tested them at four different seismic levels representative of the levels of possible earthquakes specified for nuclear plants in the United States. Seismic testing of the batteries did not cause any loss of electrical capacity. 19 refs., 29 figs., 7 tabs

Preliminary Seismic Performance Evaluation of RPS Cabinet in a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kwag, Shinyoung; Oh, Jinho; Lee, Jongmin; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

This RPS cabinet mainly provides the operators with the physical interface to monitor and handle the RPS. The objective of this paper is to perform seismic analyses and evaluate the preliminary structural integrity and seismic capacity of the RPS cabinet. For this purpose, a 3-D finite element model of the RPS cabinet is developed and its modal analyses are carried out for analyzing the dynamic characteristics. Response time history analyses and related safety evaluation are performed for the RPS cabinet subjected to seismic loads. Finally, the seismic margin and seismic fragility of the RPS cabinet are investigated. The seismic analysis, and preliminary structural integrity and seismic margin of the RPS cabinet under self weight and seismic load have been evaluated. For this purpose, 3-D finite element models of the RPS cabinet were developed. A modal analysis, response time history analysis, and seismic fragility analysis were then performed. From the structural analysis results, the RPS cabinet is below the structural design limit under PGA 0.3g (hor.) and 0.2g (ver.) and structurally withstands until PGA 3g (hor.) and 2g (ver.)

A review of procedures available to seismically requalify operating nuclear plant structures, equipment and distribution systems

International Nuclear Information System (INIS)

Stevenson, J.D.

1985-01-01

It is well known that the loads and procedures used to seismically qualify nuclear power plant structures and components have changed dramatically during the past 15 to 20 years. In this paper, the various methods available to seismically qualify or requalify structures and components in operating nuclear power plants are identified and the advantages and disadvantages of each briefly summarized. (orig.)

Seismic re-evaluation of Heavy Water Plant, Kota

International Nuclear Information System (INIS)

Parulekar, Y.M.; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

2003-10-01

This report deals with seismic re-evaluation of Heavy Water Plant, Kota. Heavy Water Plant, Kota handles considerable amount of H 2 S gas, which is very toxic. During the original design stage as per IS 1893-1966 seismic coefficient for zone-I was zero. Therefore earthquake and its effects were not considered while designing the heavy water plant structures. However as per IS 1893 (1984) the seismic coefficient for zone-I is 0.01 g. Hence seismic re-evaluation of various structures of the heavy water plant is carried out. Analysis of the heavy water plant structures was carried out for self weight, equipment load and earthquake load. Pressure loading was also considered in case of H 2 S storage tanks. Soil structure interaction effect was considered in the analysis. The combined stresses in the structures due to earthquake and dead load were checked with the allowable stresses. (author)

Introduction of conditional mean spectrum and conditional spectrum in the practice of seismic safety evaluation in China

Science.gov (United States)

Ji, Kun; Bouaanani, Najib; Wen, Ruizhi; Ren, Yefei

2018-05-01

This paper aims at implementing and introducing the use of conditional mean spectrum (CMS) and conditional spectrum (CS) as the main input parameters in the practice of seismic safety evaluation (SSE) in China, instead of the currently used uniform hazard spectrum (UHS). For this purpose, a procedure for M-R-epsilon seismic hazard deaggregation in China was first developed. For illustration purposes, two different typical sites in China, with one to two dominant seismic zones, were considered as examples to carry out seismic hazard deaggregation and illustrate the construction of CMS/CS. Two types of correlation coefficients were used to generate CMS and the results were compared over a vibration period range of interest. Ground motion records were selected from the NSMONS (2007-2015) and PEER NGA-West2 databases to correspond to the target CMS and CS. Hazard consistency of the spectral accelerations of the selected ground motion records was evaluated and validated by computing the annual exceedance probability rate of the response spectra and comparing the results to the hazard curve corresponding to each site of concern at different periods. The tools developed in this work and their illustrative application to specific case studies in China are a first step towards the adoption of CMS and CS into the practice of seismic safety evaluation in this country.

A quantitative evaluation of seismic margin of typical sodium piping

International Nuclear Information System (INIS)

Morishita, Masaki

1999-05-01

It is widely recognized that the current seismic design methods for piping involve a large amount of safety margin. From this viewpoint, a series of seismic analyses and evaluations with various design codes were made on typical LMFBR main sodium piping systems. Actual capability against seismic loads were also estimated on the piping systems. Margins contained in the current codes were quantified based on these results, and potential benefits and impacts to the piping seismic design were assessed on possible mitigation of the current code allowables. From the study, the following points were clarified; 1) A combination of inelastic time history analysis and true (without margin)strength capability allows several to twenty times as large seismic load compared with the allowable load with the current methods. 2) The new rule of the ASME is relatively compatible with the results of inelastic analysis evaluation. Hence, this new rule might be a goal for the mitigation of seismic design rule. 3) With this mitigation, seismic design accommodation such as equipping with a large number of seismic supports may become unnecessary. (author)

Seismic re-evaluation of French nuclear power plants

International Nuclear Information System (INIS)

Andrieu, R.

1995-01-01

After a presentation of the seismic inputs which have been taken into account in the design of the French Nuclear Power Plants, the re-assessed values of these inputs are shown. Some considerations about the specificity of the French PWR program with regard to the standardisation of plants are given together with the present objectives of seismic re-evaluations. Finally the main results of the seismic re-analysis being performed for the Phenix Fast Reactor are considered. (author)

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

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)

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)

Seismic hazard assessment of Iran

Directory of Open Access Journals (Sweden)

M. Ghafory-Ashtiany

1999-06-01

Full Text Available The development of the new seismic hazard map of Iran is based on probabilistic seismic hazard computation using the historical earthquakes data, geology, tectonics, fault activity and seismic source models in Iran. These maps have been prepared to indicate the earthquake hazard of Iran in the form of iso-acceleration contour lines, and seismic hazard zoning, by using current probabilistic procedures. They display the probabilistic estimates of Peak Ground Acceleration (PGA for the return periods of 75 and 475 years. The maps have been divided into intervals of 0.25 degrees in both latitudinal and longitudinal directions to calculate the peak ground acceleration values at each grid point and draw the seismic hazard curves. The results presented in this study will provide the basis for the preparation of seismic risk maps, the estimation of earthquake insurance premiums, and the preliminary site evaluation of critical facilities.

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

Evaluation of Seismic Risk of Siberia Territory

Science.gov (United States)

Seleznev, V. S.; Soloviev, V. M.; Emanov, A. F.

The outcomes of modern geophysical researches of the Geophysical Survey SB RAS, directed on study of geodynamic situation in large industrial and civil centers on the territory of Siberia with the purpose of an evaluation of seismic risk of territories and prediction of origin of extreme situations of natural and man-caused character, are pre- sented in the paper. First of all it concerns the testing and updating of a geoinformation system developed by Russian Emergency Ministry designed for calculations regarding the seismic hazard and response to distructive earthquakes. The GIS database contains the catalogues of earthquakes and faults, seismic zonation maps, vectorized city maps, information on industrial and housing fund, data on character of building and popula- tion in inhabited places etc. The geoinformation system allows to solve on a basis of probabilistic approaches the following problems: - estimating the earthquake impact, required forces, facilities and supplies for life-support of injured population; - deter- mining the consequences of failures on chemical and explosion-dangerous objects; - optimization problems on assurance technology of conduct of salvage operations. Using this computer program, the maps of earthquake risk have been constructed for several seismically dangerous regions of Siberia. These maps display the data on the probable amount of injured people and relative economic damage from an earthquake, which can occur in various sites of the territory according to the map of seismic zona- tion. The obtained maps have allowed determining places where the detailed seismo- logical observations should be arranged. Along with it on the territory of Siberia the wide-ranging investigations with use of new methods of evaluation of physical state of industrial and civil establishments (buildings and structures, hydroelectric power stations, bridges, dams, etc.), high-performance detailed electromagnetic researches of ground conditions of city

Seismic monitoring of the Creys-Malville plant - Problems raised by the seismic behaviour of a fast breeder reactor

International Nuclear Information System (INIS)

Descleve, P.; Barrau, P.

1988-01-01

CREYS-MALVILLE reached full power in December 1986 and is presently the largest sodium cooled reactor in operation. Well established procedures of safety evaluation have been used for the design but for a large size reactor special attention must be paid to the effects of seismic disturbances. This paper describes the seismic protection and monitoring system of the plant, the core behaviour which is specific to fast reactors and the test performed to verify the analyses. Finally the seismic impact on the construction can be established as an indication for future plants. (author)

Seismic evaluation of commercial plutonium fabrication plants in the United States

International Nuclear Information System (INIS)

Bernreuter, D.L.; Coats, D.W.; Murray, R.C.; Tokarz, F.J.

1979-01-01

This report is an overview of Lawrence Livermore National Laboratory's seismic assessment of six commercial plutonium fabrication plants licensed by the US Nuclear Regulatory Commission (NRC) before September 2, 1971. The seismic assessment generally has three parts: (1) documentation of the structural condition of each facility and its critical equipment; (2) characterization of the seismic hazard (i.e., determination of peak ground acceleration vs return period for each site); and (3) evaluation of seismic capacity to determine ground motion levels at which critical structures and equipment fail. The failure evaluation used structural capacities of median-centered strength characteristics of the as-built configurations from (1) and seismic hazard input from (2). Results of the assessment were partial input for an overall natural risks study by the NRC

An academic program for experience-based seismic evaluation

International Nuclear Information System (INIS)

Nix, S.J.; Meyer, W.; Clemence, S.P.

1990-01-01

The authors have been involved in a project, sponsored by the Niagara Mohawk Power Corporation, to develop knowledge-based expert systems to aid in the implementation of the Seismic Qualification Utility Group (SQUG) approach for the seismic qualification of equipment in operating nuclear power plants. This approach, being founded on the use of engineering judgment in the application of prior earthquake experience data, requires comprehensive training. There seems to be general consensus that the experience-based approach is a more cost-effective means of qualifying nuclear power plant equipment when compared to the more traditional analytical methods. The experience-based approach has a number of potential applications in civil engineering, including bridge evaluation and design, seismic adequacy of general structures, foundation design, and water and wastewater treatment plant design and operation. The objective of this paper is to outline an academic curriculum, at the master's level, to educate structural engineers to use and further develop the experience-based approach for seismic evaluation. In the long term, this could lead to the development of academic programs in experience-based assessment and design for a wide range of applications in maintaining the nation's infrastructure

Overview of seismic re-evaluation methodologies

International Nuclear Information System (INIS)

Campbell, R.D.; Johnson, J.J.

1993-01-01

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial U.S. Nuclear Power Plants and U.S. Government research reactors, production reactors and process facilities. The methodologies for resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic safety assessment criteria (SPSA). The criteria for SMAs and SPSAs have been based on realistically considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the U.S. but there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non U.S. reactors including WWER reactors. (author)

Review on Rapid Seismic Vulnerability Assessment for Bulk of Buildings

Science.gov (United States)

Nanda, R. P.; Majhi, D. R.

2013-09-01

This paper provides a brief overview of rapid visual screening (RVS) procedures available in different countries with a comparison among all the methods. Seismic evaluation guidelines from, USA, Canada, Japan, New Zealand, India, Europe, Italy, UNDP, with other methods are reviewed from the perspective of their applicability to developing countries. The review shows clearly that some of the RVS procedures are unsuited for potential use in developing countries. It is expected that this comparative assessment of various evaluation schemes will help to identify the most essential components of such a procedure for use in India and other developing countries, which is not only robust, reliable but also easy to use with available resources. It appears that Federal Emergency Management Agency (FEMA) 154 and New Zealand Draft Code approaches can be suitably combined to develop a transparent, reasonably rigorous and generalized procedure for seismic evaluation of buildings in developing countries.

A probabilistic seismic risk assessment procedure for nuclear power plants: (II) Application

Science.gov (United States)

Huang, Y.-N.; Whittaker, A.S.; Luco, N.

2011-01-01

This paper presents the procedures and results of intensity- and time-based seismic risk assessments of a sample nuclear power plant (NPP) to demonstrate the risk-assessment methodology proposed in its companion paper. The intensity-based assessments include three sets of sensitivity studies to identify the impact of the following factors on the seismic vulnerability of the sample NPP, namely: (1) the description of fragility curves for primary and secondary components of NPPs, (2) the number of simulations of NPP response required for risk assessment, and (3) the correlation in responses between NPP components. The time-based assessment is performed as a series of intensity-based assessments. The studies illustrate the utility of the response-based fragility curves and the inclusion of the correlation in the responses of NPP components directly in the risk computation. ?? 2011 Published by Elsevier B.V.

78 FR 13097 - Electric Power Research Institute; Seismic Evaluation Guidance

Science.gov (United States)

2013-02-26

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0038] Electric Power Research Institute; Seismic... Electric Power Research Institute (EPRI)-1025287, ``Seismic Evaluation Guidance: Screening, Prioritization... guidance and clarification of an acceptable approach to assist nuclear power reactor licensees when...

Evaluation of seismic characteristics and structural integrity for the cabinet of HANARO seismic monitoring analysis system

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Yoon, Doo Byung

2003-06-01

The HANARO SMAS(Seismic Monitoring Analysis System) is classified as Non-Nuclear Safety(NNS), seismic category I, and quality class T. It is required that this system can perform required functions, which are to preserve its structural integrity during and after an OBE or SSE. In this work, the structural integrity and seismic characteristics of the cabinet of the newly developed SMAS have been estimated. The most parts of the cabinet are identically designed with those of Yonggwhang and Gori Nuclear Power Plants(NPPs), unit 1 that successfully completed the required seismic qualification tests. The structure of the cabinet of the SMAS is manufactured by the manufacturer of the cabinet of Yonggwhang and Gori NPPs. To evaluate the seismic characteristics of the SMAS, the RRS(Required Response Spectra) of the newly developed cabinet are compared with those of Yonggwhang and Gori NPPs, unit 1. In addition, natural frequencies of the cabinet of HANARO, Yonggwhang, and Gori NPPs were measured for the comparison of the seismic characteristics of the installed cabinets. In case of HANARO, the bottom of the cabinet is welded to the base plate. The base plate is fixed to the concrete foundation by using anchor bolts. For the evaluation of the structural integrity of the welding parts and the anchor bolts, the maximum stresses and forces of the welding parts and the anchor bolts due to seismic loading are estimated. The analysis results show that maximum stresses and forces are less than the allowable limits. This new SMAS is operating at HANARO instrument room to acquire and analyze the signal of earthquake.

On development and improvement of evaluation techniques for seismic ground motion

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Issues regarding evaluation of active fault and ground motion for formulation of design basis ground motion (Ss) were prescribed in 'NSC seismic and tsunami safety reviewing manual' in 2012. Moreover, Nuclear Regulation Authority (NRA) is establishing the new seismic safety guideline. In this theme following four subjects were investigated to resolve the important problems for ground motion evaluation, (1) advanced evaluation of ground motion using fault model and uncertainty; (2) improving evaluation of ground motion using attenuation relation of response spectrum; (3) development of advanced and generic techniques for ground motion observation and observation tool in deep borehole; (4) improving the evaluation of site effect and seismic wave propagation characteristics. In addition as emergency requirements from NRA following two subjects were also investigated; (5) hazard evaluation development on fault displacement; (6) ground motion evaluation at near-by source location. Obtained results will be reflected not only in the domestic guideline established by NRA but in the national safety review and also in the safety standard guidelines of the International Atomic Energy Agency (IAEA) through its Extra-Budgetary Program (EBP), thereby contributing to technical cooperation in global nuclear seismic safety. (author)

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)

Re-evaluation and updating of the seismic hazard of Lebanon

Science.gov (United States)

Huijer, Carla; Harajli, Mohamed; Sadek, Salah

2016-01-01

This paper presents the results of a study undertaken to evaluate the implications of the newly mapped offshore Mount Lebanon Thrust (MLT) fault system on the seismic hazard of Lebanon and the current seismic zoning and design parameters used by the local engineering community. This re-evaluation is critical, given that the MLT is located at close proximity to the major cities and economic centers of the country. The updated seismic hazard was assessed using probabilistic methods of analysis. The potential sources of seismic activities that affect Lebanon were integrated along with any/all newly established characteristics within an updated database which includes the newly mapped fault system. The earthquake recurrence relationships of these sources were developed from instrumental seismology data, historical records, and earlier studies undertaken to evaluate the seismic hazard of neighboring countries. Maps of peak ground acceleration contours, based on 10 % probability of exceedance in 50 years (as per Uniform Building Code (UBC) 1997), as well as 0.2 and 1 s peak spectral acceleration contours, based on 2 % probability of exceedance in 50 years (as per International Building Code (IBC) 2012), were also developed. Finally, spectral charts for the main coastal cities of Beirut, Tripoli, Jounieh, Byblos, Saida, and Tyre are provided for use by designers.

Evaluation and assessment of nuclear power plant seismic methodology

International Nuclear Information System (INIS)

Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

1977-01-01

The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology

Evaluation and assessment of nuclear power plant seismic methodology

Energy Technology Data Exchange (ETDEWEB)

Bernreuter, D.; Tokarz, F.; Wight, L.; Smith, P.; Wells, J.; Barlow, R.

1977-03-01

The major emphasis of this study is to develop a methodology that can be used to assess the current methods used for assuring the seismic safety of nuclear power plants. The proposed methodology makes use of system-analysis techniques and Monte Carlo schemes. Also, in this study, we evaluate previous assessments of the current seismic-design methodology.

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

1989-01-01

A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

1989-01-01

A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

Study on the application of ambient vibration tests to evaluate the effectiveness of seismic retrofitting

Science.gov (United States)

Liang, Li; Takaaki, Ohkubo; Guang-hui, Li

2018-03-01

In recent years, earthquakes have occurred frequently, and the seismic performance of existing school buildings has become particularly important. The main method for improving the seismic resistance of existing buildings is reinforcement. However, there are few effective methods to evaluate the effect of reinforcement. Ambient vibration measurement experiments were conducted before and after seismic retrofitting using wireless measurement system and the changes of vibration characteristics were compared. The changes of acceleration response spectrum, natural periods and vibration modes indicate that the wireless vibration measurement system can be effectively applied to evaluate the effect of seismic retrofitting. The method can evaluate the effect of seismic retrofitting qualitatively, it is difficult to evaluate the effect of seismic retrofitting quantitatively at this stage.

Seismogenic zonation and seismic hazard estimates in a Southern Italy area (Northern Apulia characterised by moderate seismicity rates

Directory of Open Access Journals (Sweden)

V. Del Gaudio

2009-02-01

Full Text Available The northernmost part of Apulia, in Southern Italy, is an emerged portion of the Adriatic plate, which in past centuries was hit by at least three disastrous earthquakes and at present is occasionally affected by seismic events of moderate energy. In the latest seismic hazard assessment carried out in Italy at national scale, the adopted seismogenic zonation (named ZS9 has defined for this area a single zone including parts of different structural units (chain, foredeep, foreland. However significant seismic behaviour differences were revealed among them by our recent studies and, therefore, we re-evaluated local seismic hazard by adopting a zonation, named ZNA, modifying the ZS9 to separate areas of Northern Apulia belonging to different structural domains. To overcome the problem of the limited datasets of historical events available for small zones having a relatively low rate of earthquake recurrence, an approach was adopted that integrates historical and instrumental event data. The latter were declustered with a procedure specifically devised to process datasets of low to moderate magnitude shocks. Seismicity rates were then calculated following alternative procedural choices, according to a "logic tree" approach, to explore the influence of epistemic uncertainties on the final results and to evaluate, among these, the importance of the uncertainty in seismogenic zonation. The comparison between the results obtained using zonations ZNA and ZS9 confirms the well known "spreading effect" that the use of larger seismogenic zones has on hazard estimates. This effect can locally determine underestimates or overestimates by amounts that make necessary a careful reconsideration of seismic classification and building code application.

Comparison of evaluation guidelines for life-safety seismic hazards

International Nuclear Information System (INIS)

Wyllie, L.A.; Love, R.J.

1989-01-01

The guidelines presented in Design Evaluation guidelines for Department of Energy Facilities Subjected to natural Phenomena Hazards (UCRL 15910 Draft; May 1989) include evaluation criteria for existing Department of Energy buildings subjected to earthquakes. These criteria were developed at the Lawrence Livermore National Laboratory for use in both the seismic design of new structures and the evaluation of existing structures. ATC-14: Evaluating The Seismic Resistance of Existing Buildings developed by the Applied Technology Council, consists of guidelines and criteria for identifying the buildings or building components that present unacceptable risk to human lives. This paper compares and contrasts the two evaluation guidelines for existing buildings using a prototype building as an example. The prototype building is a seven story, concrete shear wall building assuming a General Use Occupancy

Study on seismic reliability for foundation grounds and surrounding slopes of nuclear power plants. Proposal of evaluation methodology and integration of seismic reliability evaluation system

International Nuclear Information System (INIS)

Ohtori, Yasuki; Kanatani, Mamoru

2006-01-01

This paper proposes an evaluation methodology of annual probability of failure for soil structures subjected to earthquakes and integrates the analysis system for seismic reliability of soil structures. The method is based on margin analysis, that evaluates the ground motion level at which structure is damaged. First, ground motion index that is strongly correlated with damage or response of the specific structure, is selected. The ultimate strength in terms of selected ground motion index is then evaluated. Next, variation of soil properties is taken into account for the evaluation of seismic stability of structures. The variation of the safety factor (SF) is evaluated and then the variation is converted into the variation of the specific ground motion index. Finally, the fragility curve is developed and then the annual probability of failure is evaluated combined with seismic hazard curve. The system facilitates the assessment of seismic reliability. A generator of random numbers, dynamic analysis program and stability analysis program are incorporated into one package. Once we define a structural model, distribution of the soil properties, input ground motions and so forth, list of safety factors for each sliding line is obtained. Monte Carlo Simulation (MCS), Latin Hypercube Sampling (LHS), point estimation method (PEM) and first order second moment (FOSM) implemented in this system are also introduced. As numerical examples, a ground foundation and a surrounding slope are assessed using the proposed method and the integrated system. (author)

Seismic evaluation of nuclear installations

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1997-01-01

Some considerations regarding extreme external events, natural or man-induce, such as earthquakes, floods, air crashes, etc, shall be done for nuclear facilities to minimizing the potential impact of the installation on the public and the environment. In this paper the main aspects of the seismic evaluation of nuclear facilities (except the nuclear power reactors) will be presented based on different codes and standards. (author). 7 refs., 2 tabs

Seismic capacity evaluation of a group of vertical U-tube heat exchanger with support frames for seismic PSA

International Nuclear Information System (INIS)

Watanabe, Yuichi; Muramatsu, Ken; Oikawa, Tetsukuni

2005-01-01

This paper presents an evaluation of seismic capacity of a group of vertical U-tube type heat exchangers (HXs) with support frames to contribute to refinement of seismic capacity data for seismic Probabilistic Safety Assessment (PSA) in Japan. According to usual practice of seismic PSAs, capacity of component is represented as a log-normally distributed random variable defined by a median and logarithmic standard deviations (LSDs), which represent inherent randomness about the median, β r , and uncertainty in the median due to lack of knowledge, β u . Using design specifications of four HXs for residual heat removal systems of 1100 MWe BWRs, the authors evaluated a generic capacity of HXs with a LSD for uncertainty due to lack of knowledge to take into account design variability. The median capacity was evaluated by the use of a time history response analysis with a detailed model for a selected representative HX, which was extended from a model used in seismic design. The LSD for uncertainty due to lack of knowledge was evaluated with consideration of the variabilities in three influential design parameters, i.e., diameter of anchor bolt, weight of HX and position of center of gravity of HX with the detailed model and a simplified static model. The LSD for uncertainty due to randomness was determined from the variability in material property. The dominant failure mode of HXs was identified as the failure of anchor bolts of lugs mainly due to shearing stress. The capacity expressed in terms of zero period acceleration on the foundation of HX was evaluated to be 4180 Gal (4.3 g) for median, LSD for uncertainty due to randomness was 0.11 and LSD due to lack of knowledge was 0.21-0.53 depending on combination of the variabilities in design parameters to be considered

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)

Analysis of parameter uncertainties in the assessment of seismic risk for nuclear power plants

International Nuclear Information System (INIS)

Yucemen, S.M.

1981-04-01

Probabilistic and statistical methods are used to develop a procedure by which the seismic risk at a specific site can be systematically analyzed. The proposed probabilistic procedure provides a consisted method for the modelling, analysis and updating of uncertainties that are involved in the seismic risk analysis for nuclear power plants. Methods are proposed for including these uncertainties in the final value of calculated risks. Two specific case studies are presented in detail to illustrate the application of the probabilistic method of seismic risk evaluation and to investigate the sensitivity of results to different assumptions

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

International symposium on seismic evaluation of existing nuclear facilities. Book of invited and contributed papers

International Nuclear Information System (INIS)

2003-08-01

In the past decade, seismic evaluation of existing Nuclear Power Plants (NPPs) has been an issue in western countries (particularly for the east of North America and for older NPPs in Europe) as well as in eastern European countries where systematic reviews of NPPs were carried out. Several Member States have still on-going seismic upgrading programmes. Presently. projects of plant life extension create an additional interest in safety evaluation of existing NPPs. Seismic evaluation is also an issue for other nuclear facilities. In western countries. some older facilities (laboratories, research reactors, fuel plants...) have been designed without taking into account (or poorly taking into account) earthquake input, even on seismic sites. In eastern countries, the situation is not yet clearly evaluated. In several countries concerns are expressed regarding research reactors. Generally speaking, the seismic evaluation of these nuclear facilities is not so advanced as the evaluation of NPPs and presents a wider range of different situations. Those safety issues raised by the seismic evaluation of existing NPPs are addressed in an IAEA Safety Report 'Seismic Evaluation of Existing Nuclear Power Plants' (2003). Other nuclear facilities are, at least partly, covered by IAEA-TECDOC 1347 'Design of Nuclear Facilities other than NPPs in Relation to External Events, with a Special Emphasis on Earthquake' (2003) that supersedes the former TECDOC 348 'Earthquake Resistant Design of Nuclear Facilities with Limited Radioactive Inventory' (1985). Concurrently with the publishing of relevant documentation, the IAEA has organized this Symposium in order to foster the exchange of information on topical issues in seismic evaluation of existing nuclear facilities, with the aim of: consolidating an international consensus on the present status of these issues; promoting a homogeneous engineering approach of their resolution; identifying the needs for strengthening international co

International symposium on seismic evaluation of existing nuclear facilities. Book of invited and contributed papers

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-08-01

In the past decade, seismic evaluation of existing Nuclear Power Plants (NPPs) has been an issue in western countries (particularly for the east of North America and for older NPPs in Europe) as well as in eastern European countries where systematic reviews of NPPs were carried out. Several Member States have still on-going seismic upgrading programmes. Presently. projects of plant life extension create an additional interest in safety evaluation of existing NPPs. Seismic evaluation is also an issue for other nuclear facilities. In western countries. some older facilities (laboratories, research reactors, fuel plants...) have been designed without taking into account (or poorly taking into account) earthquake input, even on seismic sites. In eastern countries, the situation is not yet clearly evaluated. In several countries concerns are expressed regarding research reactors. Generally speaking, the seismic evaluation of these nuclear facilities is not so advanced as the evaluation of NPPs and presents a wider range of different situations. Those safety issues raised by the seismic evaluation of existing NPPs are addressed in an IAEA Safety Report 'Seismic Evaluation of Existing Nuclear Power Plants' (2003). Other nuclear facilities are, at least partly, covered by IAEA-TECDOC 1347 'Design of Nuclear Facilities other than NPPs in Relation to External Events, with a Special Emphasis on Earthquake' (2003) that supersedes the former TECDOC 348 'Earthquake Resistant Design of Nuclear Facilities with Limited Radioactive Inventory' (1985). Concurrently with the publishing of relevant documentation, the IAEA has organized this Symposium in order to foster the exchange of information on topical issues in seismic evaluation of existing nuclear facilities, with the aim of: consolidating an international consensus on the present status of these issues; promoting a homogeneous engineering approach of their resolution; identifying the needs for strengthening international co

Use of the Delphi approach in seismic qualification of existing electrical and mechanical equipment and distribution systems

International Nuclear Information System (INIS)

Stevenson, J.D.

1983-01-01

In this paper a method of seismic reevaluation is described which permits every seismic Category I component and distribution system to be evaluated and an explicit estimate of seismic capability, in terms of an acceleration level which defines limits of structural adequacy and leak-tight integrity, be defined for each item. The procedure employees the Delphi method where a team of seismic design experts independently inspect each component and distribution system in its as installed condition in the field and estimate the acceleration level that component or system could withstand and still meet applicable seismic design code or other criteria requirements. The accuracy and any potential bias in the estimates can be evaluated by a small control sample where selected components which were also surveyed by the Delphi team are independently evaluated in detail using current analytical techniques. This entire procedure can be accomplished for an estimated 5000 to 10,000 manhours per plant

Expert system GIP-WWER for verification of seismic adequacy of WWER equipment

International Nuclear Information System (INIS)

Masopust, R.

1999-01-01

The aim of this report is to describe the modified Generic Implementation Procedure (GIP) titled GIP-WWER which can be used to verify seismic adequacy of the safe shutdown mechanical and electrical equipment and distribution systems of operating or constructed WWER NPPs, namely WWER-440/213 type. The WWER-GIP procedure was prepared using available information contained in GIP and the experience taken from various seismic inspections and evaluations of WWER type NPPs performed in the last five years

Seismic margins review of nuclear power plants: Fragility aspects

International Nuclear Information System (INIS)

Ravindra, M.K.; Hardy, G.S.; Hashimoto, P.S.

1987-01-01

The fragility analysis is utilised in the seismic margin review in initial screening of certain components in the plant based on their generically high seismic capacities. A detailed walkdown of the plant is conducted to confirm that the initial screening is valid i.e., the generically high seismic capacity components do not possess any potential weaknesses (e.g., inadequate bracing, inadequate anchorage and potential systems interaction). For the components that are screened in, their seismic capacities are evaluated using either a probabilistic analysis of a deterministic evaluation. Based on a system analysis, the Boolean expressions for critical accident sequences are derived. These Boolean expressions are quantified using the component fragilities and nonseismic unavailabilities of components. The final product is the High Confidence Low Probability of Failure (HCLPF) capacity of the plant and the identification of potential seismic vulnerabilities in the plant. The objective of the paper is to describe the application of fragility analysis procedures in the seismic margin review of Maine Yankee and to document the insights obtained in this trial plant review. (orig./HP)

Analysis of the seismic response of a fast reactor core

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1984-01-01

This report deals with the methods to apply for a correct evaluation of the reactor core seismic response. Reference is made to up-to-date design data concerning the PEC core, taking into account the presence of the core-restraint plate located close to the PEC core elements top and applying the optimized iterative procedure between the vessel linear calculation and the non-linear ones limited to the core, which had been described in a previous report. It is demonstrated that the convergence of this procedure is very fast, similar to what obtained in the calculations of the cited report, carried out with preliminary data, and it is shown that the cited methods allow a reliable evaluation of the excitation time histories for the experimental tests in support of the seismic verification of the shutdown system and the core of a fast reactor, as well as relevant data for the experimental, structural and functional, verification of the core elements in the case of seismic loads

Seismic re-evaluation of piping systems of heavy water plant, Kota

International Nuclear Information System (INIS)

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

2002-05-01

Heavy Water Plant, Kota is the first indigenous heavy water plant built in India. The plant started operation in the year 1985 and it is approaching the completion of its originally stipulated design life. In view of the excellent record of plant operation for the past so many years, it has been planned to carry out various exercises for the life extension of the plant. In the first stage, evaluation of operation stresses was carried out for the process critical piping layouts and equipment, which are connected with 25 process critical nozzle locations, identified based on past history of the plant performance. Fatigue life evaluation has been carried out to fmd out the Cumulative Usage Factor, which helps in arriving at a decision regarding the life extension of the plant. The results of these exercises have been already reported separately vide BARC/200I /E/O04. In the second stage, seismic reevaluation of the plant has been carried out to assess its ability to maintain its integ:rity in case of a seismic event. The aim of this exercise is to assess the effects of the maximum probable earthquake at the plant site on the various systems and components of the plant. This exercise is further aimed at ensuring the adequacy of seismic supports to maintain the integrity of the system in case of a seismic event and to suggest some retrofitting measures, if required. Seismic re-evaluation of the piping of Heavy Water Plant, Kota has been performed taking into account the interaction effects from the connected equipment. Each layout has been qualified using the latest provisions of ASME Code Section III, Subsection ND wherein the earthquake loading has been considered as a reversing dynamic load. The maximum combined stresses for all the layouts due to pressure, weight and seismic loadings have been found to be well within the code allowable limit. Therefore, it has been concluded that during a maximum probable seismic event, the possibility of pipe rupture can be safely

Seismic Vulnerability Evaluation of a Three-Span Continuous Beam Railway Bridge

Directory of Open Access Journals (Sweden)

Chongwen Jiang

2017-01-01

Full Text Available In order to evaluate the seismic vulnerability of a railway bridge, a nonlinear finite element model of typical three-span continuous beam bridge on the Sichuan-Tibet railway in China was built. It further aimed at performing a probabilistic seismic demand analysis based on the seismic performance of the above-mentioned bridge. Firstly, the uncertainties of bridge parameters were analyzed while a set of finite element model samples were formulated with Latin hypercube sampling method. Secondly, under Wenchuan earthquake ground motions, an incremental dynamic method (IDA analysis was performed, and the seismic peak responses of bridge components were recorded. Thirdly, the probabilistic seismic demand model for the bridge principal components under the prerequisite of two different kinds of bearing, with and without seismic isolation, was generated. Finally, comparison was drawn to further ascertain the effect of two different kinds of bearings on the fragility components. Based on the reliability theory, results were presented concerning the seismic fragility curves.

Regulatory application of seismic experience data for nuclear power plants in the United States

Energy Technology Data Exchange (ETDEWEB)

Chen, Pei-Ying [Nuclear Regulatory Commission, Washington, DC (United States)

1997-03-01

On the basis of its review and evaluation (Reference 3) of the SQUG GIP (Reference 2) and on the basis of the differences between current seismic qualification requirements and the criteria and procedures provided in the GIP, the NRC staff does not consider the USI A-46 methodology given in the GIP to be a `seismic qualification` procedure. Rather, the staff considers the GIP methodology to be a seismic adequacy verification procedure, which was developed on the basis of generic equipment earthquake experience data, supplemented by generic equipment test data. The implementation of the GIP approach for USI A-46 plants provides safety enhancement, in certain aspects, beyond the original licensing bases. Therefore, the GIP methodology is an acceptable evaluation method, for USI A-46 plants only, to verify the seismic adequacy of the safe-shutdown equipment installed in the NPPs in the United States. With the new development in the experience-based approach for seismic qualification of equipment currently underway in the U.S. nuclear industry, there is a potential for future regulatory application of an experience-based approach as a seismic qualification method for certain selected equipment installed in NPPs in the United States. However, industry`s use of the experience-based approach will be dependent on the submittal and staff approval of this approach. (J.P.N.)

Delineation of seismic source zones based on seismicity parameters ...

Indian Academy of Sciences (India)

In the present study, an attempt has been made to delineate seismic source zones in the study area (south India) based on the seismicity parameters. Seismicity parameters and the maximum probable earthquake for these source zones were evaluated and were used in the hazard evaluation. The probabilistic evaluation of ...

Evaluation of induced seismicity forecast models in the Induced Seismicity Test Bench

Science.gov (United States)

Király, Eszter; Gischig, Valentin; Zechar, Jeremy; Doetsch, Joseph; Karvounis, Dimitrios; Wiemer, Stefan

2016-04-01

Induced earthquakes often accompany fluid injection, and the seismic hazard they pose threatens various underground engineering projects. Models to monitor and control induced seismic hazard with traffic light systems should be probabilistic, forward-looking, and updated as new data arrive. Here, we propose an Induced Seismicity Test Bench to test and rank such models. We apply the test bench to data from the Basel 2006 and Soultz-sous-Forêts 2004 geothermal stimulation projects, and we assess forecasts from two models that incorporate a different mix of physical understanding and stochastic representation of the induced sequences: Shapiro in Space (SiS) and Hydraulics and Seismics (HySei). SiS is based on three pillars: the seismicity rate is computed with help of the seismogenic index and a simple exponential decay of the seismicity; the magnitude distribution follows the Gutenberg-Richter relation; and seismicity is distributed in space based on smoothing seismicity during the learning period with 3D Gaussian kernels. The HySei model describes seismicity triggered by pressure diffusion with irreversible permeability enhancement. Our results show that neither model is fully superior to the other. HySei forecasts the seismicity rate well, but is only mediocre at forecasting the spatial distribution. On the other hand, SiS forecasts the spatial distribution well but not the seismicity rate. The shut-in phase is a difficult moment for both models in both reservoirs: the models tend to underpredict the seismicity rate around, and shortly after, shut-in. Ensemble models that combine HySei's rate forecast with SiS's spatial forecast outperform each individual model.

Seismic effects on technological equipment and systems of nuclear power plants

International Nuclear Information System (INIS)

Masopust, R.; Pecinka, L.; Podrouzek, J.

1983-01-01

A survey is given of problems related to the construction of nuclear power plants with regard to seismic resistance. Sei--smic resistance of technological equipment is evaluated by experimental trials, calculation or the combination of both. Existing and future standards are given for the given field. The Czechoslovak situation is discussed as related to the construction of the Mochovce nuclear power plant. Procedures for testing seismic resistance, types of tests and methods of simulating seismic excitation are described. Antiseismic measures together with structural elements for limiting the seismic effects on technological equipment and nuclear power plant systems are summed up on the basis of foreign experience. (E.F.)

Integrated software system for seismic evaluation of nuclear power plant structures

International Nuclear Information System (INIS)

Xu, J.; Graves, H.L.

1993-01-01

The computer software CARES (Computer Analysis for Rapid Evaluation of Structures) was developed by the Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission. It represents an effort to utilize established numerical methodologies commonly employed by industry for structural safety evaluations of nuclear power plant facilities and incorporates them into an integrated computer software package operated on personal computers. CARES was developed with the objective of including all aspects of seismic performance evaluation of nuclear power structures. It can be used to evaluate the validity and accuracy of analysis methodologies used for structural safety evaluations of nuclear power plants by various utilities. CARES has a modular format, each module performing a specific type of analysis. The seismic module integrates all the steps of a complete seismic analysis into a single package with many user-friendly features such as interactiveness and quick turnaround. Linear structural theory and pseudo-linear convolution theory are utilized as the bases for the development with a special emphasis on the nuclear regulatory requirements for structural safety of nuclear plants. The organization of the seismic module is arranged in eight options, each performing a specific step of the analysis with most of input/output interfacing processed by the general manager. Finally, CARES provides comprehensive post-processing capability for displaying results graphically or in tabular form so that direct comparisons can be easily made. (author)

Evaluation of Seismic Behavior of Steel Braced Frames with Controlled Rocking System and Energy Dissipating Fuses

Directory of Open Access Journals (Sweden)

Hassan Amirzehni

2016-12-01

Full Text Available The self-centering rocking steel braced frames are new type of seismic lateral-force resisting systems that are developed with aim to limiting structural damages, minimizing residual drifts on systems and creating easy and inexpensive reconstruction capability, after sever earthquakes. In Steel braced frames with controlled rocking system, column bases on seismic resisting frame are not attached to the foundation and the frame allowed to rock freely. The task of restoring the rotated frame to its initial location is on post-tensioned cables, which attaches top of the frame to foundation. The design of post tensioned stands and braced frame members is such that during earthquakes they remain in elastic region. Seismic energy, dissipates by plastic deformations in replaceable elements on each rock of frame. In current research work, the seismic behavior of this type of lateral resisting systems is evaluated. The research conducted on a one bay steel braced frame with controlled rocking system that is analyzed using nonlinear dynamic time history analysis (NLTHA procedure. The frame is subjected to JMA-Kobe and Northridge ground motions records that are scaled to unit, 1.2 and 1.5 times of maximum considered earthquake (MCE ground motion level intensity. Extracted results show that seismic behavior of this type of lateral force resisting systems are so desirable even under MCE ground motion levels. The only anxiety is about occurring fatigue in post-tensioned strands that endangers overall stability of system.

Methodological approach for the seismic backfitting of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Galli, P.; Muzzi, F.; Ruggieri, G.; Zola, M.

1993-01-01

In the frame of the assessment of the seismic adequacy of the operating Nuclear Power Plants in East Europe, the main problem to match with is the difficulty to work about already existing plants. Moreover consolidated standards and procedures for seismic design, verification and qualification exist for new structures and equipment, then the extension to operating plants requires a lot of engineering judgement. The paper highlights the importance of: identification of seismic safety related systems and components; site specific seismic input definition in agreement with international standards; computation of seismic loads accounting for soil-structure interaction and appropriate structural modelling; overall stability verification of the plant (soil bearing capacity, soil liquefaction, sliding, overturning); ductility effects in evaluation of seismic protection; engineering process for the qualification of components and systems and walkdown procedures and identification of remedial measures (easy fixes and complex fixes). Some examples are reported referred to the more recent ISMES activities in the field

Evaluation of seismic induced CDF and ΔCDF with considering the uncertainty reduction research results

International Nuclear Information System (INIS)

Hahm, Daegi; Choi, In Kil

2012-01-01

In the seismic probabilistic safety assessment (SPSA) of nuclear power plants (NPPs), the efficient and rational methodology to dealing the uncertainty factors are required to increase the reliability of the SPSA results. To reduce the uncertainties in the SPSA approach, many research activities were performed by Korea Atomic Energy Research Institute (KAERI) during the last 5 years mid and long term nuclear research and development program of the ministry of education, science and technology. These outcomes can be implemented to the update or reevaluation of previous NPP's SPSA results. In this study, we applied these uncertainty reduction research results to the update of the SPSA procedure of the target reference plant, i.e., Ulchin unit 5/6 NPP. The refined topics from the SPSA procedure are the seismic fragility, the seismic hazard, and the risk quantification. The detailed process and results are described in the next sections

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

Seismic evaluation of the Mors Dome

International Nuclear Information System (INIS)

Kreitz, E.

1982-01-01

The ''Seismic Case History'' of the Mors saltdome was already published in detail by ELSAM/ELKRAFT so only a few important points need to be mentioned here: (a) Processing and interpretation of the seismic material. (b) Stratigraphic classification of the most important seismic reflection horizons. (c) Construction of the depth sections and description of the saltdome model. (d) Investigations of the problematic salt overhang using interactive seismic modelling. (EG)

Development of seismic PSA methodology at JAERI

International Nuclear Information System (INIS)

Muramatsu, K.; Ebisawa, K.; Matsumoto, K.; Oikawa, T.; Kondo, M.

1995-01-01

The Japan Atomic Energy Research Institute (JAERI) is developing a methodology for seismic probabilistic safety assessment (PSA) of nuclear power plants, aiming at providing a set of procedures, computer codes and data suitable for performing seismic PSA in Japan. In order to demonstrate the usefulness of JAERI's methodology and to obtain better understanding on the controlling factors of the results of seismic PSAs, a seismic PSA for a BWR is in progress. In the course of this PSA, various improvements were made on the methodology. In the area of the hazard analysis, the application of the current method to the model plant site is being carried out. In the area of response analysis, the response factor method was modified to consider the non-linear response effect of the building. As for the capacity evaluation of components, since capacity data for PSA in Japan are very scarce, capacities of selected components used in Japan were evaluated. In the systems analysis, the improvement of the SECOM2 code was made to perform importance analysis and sensitivity analysis for the effect of correlation of responses and correlation of capacities. This paper summarizes the recent progress of the seismic PSA research at JAERI with emphasis on the evaluation of component capacity and the methodology improvement of systems reliability analysis. (author)

A progressive methodology for seismic safety evaluation of gravity dams

International Nuclear Information System (INIS)

Ghrib, F.; Leger, P.; Tinawi, R.; Lupien, R.; Veilleux, M.

1995-01-01

A progressive methodology for the seismic safety evaluation of existing concrete gravity dams was described. The methodology was based on five structural analysis levels with increasing complexity to represent inertia forces, dam-foundation and dam-interaction mechanisms, as well as concrete cracking. The five levels were (1) preliminary screening, (2) pseudo-static method, (3) pseudo-dynamic method, (4) linear time history analysis, and (5) non-linear history analysis. The first four levels of analysis were applied for the seismic safety evaluation of Paugan gravity dam (Quebec). Results showed that internal forces from pseudo-dynamic, response spectra and transient finite element analyses could be used to interpret the dynamic stability of dams from familiar strength-based criteria. However, as soon as the base was cracked, the seismically induced forces were modified, and level IV analyses proved more suitable to handle rationally these complexities. 8 refs., 7 figs., 1 tab

Seismic test for safety evaluation of low level radioactive wastes containers

International Nuclear Information System (INIS)

Ohoka, Makoto; Horikiri, Morito

1998-08-01

Seismic safety of three-piled container system used in Tokai reprocessing center was confirmed by seismic test and computational analysis. Two types of container were evaluated, for low level noninflammable radioactive solid wastes, and for used filters wrapped by large plastic bags. Seismic integrity of three-piled containers was confirmed by evaluating response characteristics such as acceleration and displacement under the design earthquake condition S1, which is the maximum earthquake expected at the stored site during the storage time. Computational dynamic analysis was also performed, and several conclusions described below were made. (1) Response characteristics of the bottom board and the side board were different. The number of pile did not affect the response characteristics of the bottom board of each container. They behaved as a rigid body. (2) The response of the side board was larger than that of the bottom board. (3) The response depended on the direction in each board, either side or bottom. The response acceleration became larger to the seismic wave perpendicular to the plane which has the entrance for fork lift and the radioactive warning mark. (4) The maximum horizontal response displacement under the S1 seismic wave was approximately 10 mm. It is so small that it does not affect the seismic safety. (5) The stoppers to prevent fall down had no influence to the response acceleration. (6) There was no fall down to the S1 seismic wave and 2 times of S1 seismic wave, which was the maximum input condition of the test. (7) The response of the bottom board of the containers, which are main elements of fall down, had good agreements both in the test and in the computational analysis. (author)

Evaluation of seismic hazard at the northwestern part of Egypt

Science.gov (United States)

Ezzelarab, M.; Shokry, M. M. F.; Mohamed, A. M. E.; Helal, A. M. A.; Mohamed, Abuoelela A.; El-Hadidy, M. S.

2016-01-01

The objective of this study is to evaluate the seismic hazard at the northwestern Egypt using the probabilistic seismic hazard assessment approach. The Probabilistic approach was carried out based on a recent data set to take into account the historic seismicity and updated instrumental seismicity. A homogenous earthquake catalogue was compiled and a proposed seismic sources model was presented. The doubly-truncated exponential model was adopted for calculations of the recurrence parameters. Ground-motion prediction equations that recently recommended by experts and developed based upon earthquake data obtained from tectonic environments similar to those in and around the studied area were weighted and used for assessment of seismic hazard in the frame of logic tree approach. Considering a grid of 0.2° × 0.2° covering the study area, seismic hazard curves for every node were calculated. Hazard maps at bedrock conditions were produced for peak ground acceleration, in addition to six spectral periods (0.1, 0.2, 0.3, 1.0, 2.0 and 3.0 s) for return periods of 72, 475 and 2475 years. The unified hazard spectra of two selected rock sites at Alexandria and Mersa Matruh Cities were provided. Finally, the hazard curves were de-aggregated to determine the sources that contribute most of hazard level of 10% probability of exceedance in 50 years for the mentioned selected sites.

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%

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.

The Contribution of Palaeoseismology to Seismic Hazard Assessment in Site Evaluation for Nuclear Installations

International Nuclear Information System (INIS)

2015-06-01

IAEA Safety Standards Series No. SSG-9, Seismic Hazards in Site Evaluation for Nuclear Installations, published in 2010, covers all aspects of site evaluation relating to seismic hazards and recommends the use of prehistoric, historical and instrumental earthquake data in seismic hazard assessments. Prehistoric data on earthquakes cover a much longer period than do historical and instrumental data. However, gathering such data is generally difficult in most regions of the world, owing to an absence of human records. Prehistoric data on earthquakes can be obtained through the use of palaeoseismic techniques. This publication describes the current status and practices of palaeoseismology, in order to support Member States in meeting the recommendations of SSG-9 and in establishing the necessary earthquake related database for seismic hazard assessment and reassessment. At a donors’ meeting of the International Seismic Safety Centre Extrabudgetary Project in January 2011, it was suggested to develop detailed guidelines on seismic hazards. Soon after the meeting, the disastrous Great East Japan Earthquake and Tsunami of 11 March 2011 and the consequent accident at the Fukushima Daiichi nuclear power plant occurred. The importance of palaeoseismology for seismic hazard assessment in site evaluation was highlighted by the lessons learned from the Fukushima Daiichi nuclear power plant accident. However, no methodology for performing investigations using palaeoseismic techniques has so far been available in an IAEA publication. The detailed guidelines and practical tools provided here will be of value to nuclear power plant operating organizations, regulatory bodies, vendors, technical support organizations and researchers in the area of seismic hazard assessment in site evaluation for nuclear installations, and the information will be of importance in support of hazard assessments in the future

Seismic fragilities for nuclear power plant risk studies

International Nuclear Information System (INIS)

Kennedy, R.P.; Ravindra, M.K.

1983-01-01

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors

Earthquake experience and seismic qualification by indirect methods in nuclear installations

International Nuclear Information System (INIS)

2003-01-01

In recent years, many operational nuclear power plants around the world have conducted seismic re-evaluation programmes either as part of a review of seismic hazards or to comply with best international nuclear safety practices. In this connection, Member States have called on the IAEA to carry out several seismic review missions at their plants, primarily those of WWER and RBMK design. One of the critical safety issues that arose during these missions was that of seismic qualification (determination of fitness for service) of already installed plant distribution systems, equipment and components. The qualification of new components, equipment and distribution systems cannot be replicated for equipment that is already installed and operational in plants, as this process is neither feasible nor appropriate. For this reason, seismic safety experts have developed new procedures for the qualification of installed equipment: these procedures seek to demonstrate that installed equipment, through a process of comparison with new equipment, is apt for service. However, these procedures require large sets of criteria and qualification databases and call for the use of engineering judgement and experience, all of which open the door to wide margins of interpretation. In order to guarantee a sound technical basis for the qualification of in-plant equipment, currently applied to 70% to 80% of all plant equipment, the regulatory review of this type of qualification process calls for a detailed assessment of the technical procedures applied. Such an assessment is the first step towards eliminating the risk of large differences in qualification results between different plants, operators and countries, and guaranteeing the reliability of seismic re-evaluation programmes. Bearing this in mind, in 1999, the IAEA convened a seminar and technical meeting on seismic qualification under the auspices of the IAEA Technical Co-operation programme. Altogether 66 senior experts attended the

Guideline for the seismic technical evaluation of replacement items for nuclear power plants

International Nuclear Information System (INIS)

Harris, S.P.; Cushing, R.W.; Johnson, H.W.; Abeles, J.M.

1993-02-01

Seismic qualification for equipment originally installed in nuclear power plants was typically performed by the original equipment suppliers or manufactures (OES/OEM). Many of the OES/OEM no longer maintain quality assurance programs with adequate controls for supplying nuclear equipment. Utilities themselves must provide reasonable assurance in the continued seismic adequacy of such replacement items. This guideline provides practical, cost-effective techniques which can be used to provide reasonable assurance that replacement items will meet seismic performance requirements necessary to maintain the seismic design basis of commercial nuclear power plants. It also provides a method for determining when a seismic technical evaluation of replacement items (STERI) is required as part of the procurement process for spare and replacement items. Guidance on supplier program requirements necessary to maintain continued seismic adequacy and on documentation of maintaining required seismic adequacy is also included

Seismic scoping evaluation of high level liquid waste tank vaults at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Hashimoto, P.S.; Uldrich, E.D.; McGee, W.D.

1991-01-01

A seismic scoping evaluation of buried vaults enclosing high level liquid waste storage tanks at the Idaho Chemical Processing Plant has been performed. The objective of this evaluation was to scope out which of the vaults could be demonstrated to be seismically adequate against the Safe Shutdown Earthquake (SSE). Using approximate analytical methods, earthquake experience data, and engineering judgement, this study determined that one vault configuration would be expected to meet ICPP seismic design criteria, one would not be considered seismically adequate against the SSE, and one could be shown to be seismically adequate against the SSE using nonlinear analysis

A methodology for the quantitative risk assessment of major accidents triggered by seismic events

International Nuclear Information System (INIS)

Antonioni, Giacomo; Spadoni, Gigliola; Cozzani, Valerio

2007-01-01

A procedure for the quantitative risk assessment of accidents triggered by seismic events in industrial facilities was developed. The starting point of the procedure was the use of available historical data to assess the expected frequencies and the severity of seismic events. Available equipment-dependant failure probability models (vulnerability or fragility curves) were used to assess the damage probability of equipment items due to a seismic event. An analytic procedure was subsequently developed to identify, evaluate the credibility and finally assess the expected consequences of all the possible scenarios that may follow the seismic events. The procedure was implemented in a GIS-based software tool in order to manage the high number of event sequences that are likely to be generated in large industrial facilities. The developed methodology requires a limited amount of additional data with respect to those used in a conventional QRA, and yields with a limited effort a preliminary quantitative assessment of the contribution of the scenarios triggered by earthquakes to the individual and societal risk indexes. The application of the methodology to several case-studies evidenced that the scenarios initiated by seismic events may have a relevant influence on industrial risk, both raising the overall expected frequency of single scenarios and causing specific severe scenarios simultaneously involving several plant units

Seismic analysis and structure capacity evaluation of the Belene nuclear power plant

International Nuclear Information System (INIS)

Johnson, J.J.; Hashimoto, P.S.; Campbell, R.D.; Baltus, R.S.

1993-01-01

The seismic analysis and structure capacity evaluation of the Belene Nuclear Power Plant, a two-unit WWER 1000, was performed. The principal objective of the study was to review the major aspects of the seismic design including ground motion specification, foundation concept and materials, and the Unit I main reactor building structure response and capacity. The main reactor building structure /foundation/soil were modeled and analyzed by a substructure approach to soil-structure interaction (SSI) analysis. The elements of the substructure approach, implemented in the family of computer programs CLASSI, are: Specification of the free-field ground motion; Modeling the soil profile; SSI parameters; Modeling the structure; SSI-response analyses. Each of these aspects is discussed. The Belene Unit 1 main reactor building structure was evaluated to verify the seismic design with respect to current western criteria. The structural capacity evaluation included criteria development, element load distribution analysis, structural element selection, and structural element capacity evaluation. Equipment and commodity design criteria were similarly reviewed and evaluated. Methodology results and recommendations are presented. (author)

SQUG cable tray and conduit evaluation procedure

International Nuclear Information System (INIS)

Smith, P.D.; Eder, S.J.; Conoscente, J.P.

1990-01-01

Cable tray and conduit systems for electrical cables are a common feature and industrial facilities. They have an excellent performance history in past strong earthquake, even though they are rarely designed for earthquakes. Considerable data have been gathered on their performance in earthquakes and in shake table testing. The data have been used to develop a procedure for the verification of the seismic adequacy of cable tray and conduit systems in operating nuclear plants. The procedure is discussed in this paper. It will result in substantial savings, such as reduced engineering effort, fewer modifications of existing hardware, and simpler documentation, relative to alternate procedures like dynamic analysis of shake table testing. The procedure ensures safety-function in a unique manner since the methodology used to develop it (1) is based on a large body of historical data and (2) uses a relative approach of ensuring that nuclear plant systems will perform at least as well as systems that performed well in past earthquakes. (orig.)

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)

Probabilistic seismic hazards: Guidelines and constraints in evaluating results

International Nuclear Information System (INIS)

Sadigh, R.K.; Power, M.S.

1989-01-01

In conducting probabilistic seismic hazard analyses, consideration of the dispersion as well as the upper bounds on ground motion is of great significance. In particular, the truncation of ground motion levels at some upper limit would have a major influence on the computed hazard at the low-to-very-low probability levels. Additionally, other deterministic guidelines and constraints should be considered in evaluating the probabilistic seismic hazard results. In contrast to probabilistic seismic hazard evaluations, mean plus one standard deviation ground motions are typically used for deterministic estimates of ground motions from maximum events that may affect a structure. To be consistent with standard deterministic maximum estimates of ground motions values should be the highest level considered for the site. These maximum values should be associated with the largest possible event occurring at the site. Furthermore, the relationships between the ground motion level and probability of exceedance should reflect a transition from purely probabilistic assessments of ground motion at high probability levels where there are multiple chances for events to a deterministic upper bound ground motion at very low probability levels where there is very limited opportunity for maximum events to occur. In Interplate Regions, where the seismic sources may be characterized by a high-to-very-high rate of activity, the deterministic bounds will be approached or exceeded by the computer probabilistic hazard values at annual probability of exceedance levels typically as high as 10 -2 to 10 -3 . Thus, at these or lower values probability levels, probabilistically computed hazard values could be readily interpreted in the light of the deterministic constraints

Use of seismic experience data for technical evaluation of commercial grade replacement items

International Nuclear Information System (INIS)

Cushing, R.W.; Campbell, R.D.

1991-01-01

This paper reports that the purchase of commercial grade replacement items which are dedicated for use in nuclear safety related systems requires a technical evaluation for those items which cannot be purchased as like-for-like replacements. One of the requirements for performing a technical evaluation is verification of seismic adequacy. Studying the performance and failure modes of equipment which has been subjected to strong motion earthquakes has provided the industry with the ability to predict the behavior characteristics of many types of commonly replaced components and parts. The use of Seismic Experience Data has developed into an efficient means of satisfying the seismic adequacy requirement

A framework of risk-informed seismic safety evaluation of nuclear power plants in Japan

International Nuclear Information System (INIS)

Kondo, S.; Sakagami, M.; Hirano, M.; Shiba, M.

2001-01-01

A framework of risk-informed seismic design and safety evaluation of nuclear power plants is under consideration in Japan so as to utilize the progress in the seismic probabilistic safety assessment methodology. Issues resolved to introduce this framework are discussed after the concept, evaluation process and characteristics of the framework are described. (author)

Evaluation of seismic behavior of a braced tubular steel structure by pseudodynamic testing

International Nuclear Information System (INIS)

Shiny, P.B.; Javadian-Gilani, A.S.; Mahin, S.A.

1984-01-01

The inelastic seismic behavior of an X-braced, tubular steel frame is studied experimentally by means of pseudodynamic testing. The pseudodynamic method, which utilizes a numerical algorithm in the on-line computer control of a test specimen, can realistically simulate the seismic response of a structural model. This paper presents a brief outline of the experimental procedure and the results of the tubular frame tests, including the global responses, the inelastic energy-dissipation capabilities, and the failure mechanism of the frame at various excitation levels. Correlation of these results with previous experimental studies illustrates the feasibility and accuracy of the new test method

Seismic fragility analyses

International Nuclear Information System (INIS)

Kostov, Marin

2000-01-01

In the last two decades there is increasing number of probabilistic seismic risk assessments performed. The basic ideas of the procedure for performing a Probabilistic Safety Analysis (PSA) of critical structures (NUREG/CR-2300, 1983) could be used also for normal industrial and residential buildings, dams or other structures. The general formulation of the risk assessment procedure applied in this investigation is presented in Franzini, et al., 1984. The probability of failure of a structure for an expected lifetime (for example 50 years) can be obtained from the annual frequency of failure, β E determined by the relation: β E ∫[d[β(x)]/dx]P(flx)dx. β(x) is the annual frequency of exceedance of load level x (for example, the variable x may be peak ground acceleration), P(fI x) is the conditional probability of structure failure at a given seismic load level x. The problem leads to the assessment of the seismic hazard β(x) and the fragility P(fl x). The seismic hazard curves are obtained by the probabilistic seismic hazard analysis. The fragility curves are obtained after the response of the structure is defined as probabilistic and its capacity and the associated uncertainties are assessed. Finally the fragility curves are combined with the seismic loading to estimate the frequency of failure for each critical scenario. The frequency of failure due to seismic event is presented by the scenario with the highest frequency. The tools usually applied for probabilistic safety analyses of critical structures could relatively easily be adopted to ordinary structures. The key problems are the seismic hazard definitions and the fragility analyses. The fragility could be derived either based on scaling procedures or on the base of generation. Both approaches have been presented in the paper. After the seismic risk (in terms of failure probability) is assessed there are several approaches for risk reduction. Generally the methods could be classified in two groups. The

Seismic safety margins research program. Phase I final report - Overview

International Nuclear Information System (INIS)

Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Chuang, T.Y.; Cummings, G.E.; Johnson, J.J.; Mensing, R.W.; Wells, J.E.

1981-04-01

The Seismic Safety Margins Research Program (SSMRP) is a multiyear, multiphase program whose overall objective is to develop improved methods for seismic safety assessments of nuclear power plants, using a probabilistic computational procedure. The program is being carried out at the Lawrence Livermore National Laboratory and is sponsored by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Phase I of the SSMRP was successfully completed in January 1981: A probabilistic computational procedure for the seismic risk assessment of nuclear power plants has been developed and demonstrated. The methodology is implemented by three computer programs: HAZARD, which assesses the seismic hazard at a given site, SMACS, which computes in-structure and subsystem seismic responses, and SEISIM, which calculates system failure probabilities and radioactive release probabilities, given (1) the response results of SMACS, (2) a set of event trees, (3) a family of fault trees, (4) a set of structural and component fragility descriptions, and (5) a curve describing the local seismic hazard. The practicality of this methodology was demonstrated by computing preliminary release probabilities for Unit 1 of the Zion Nuclear Power Plant north of Chicago, Illinois. Studies have begun aimed at quantifying the sources of uncertainty in these computations. Numerous side studies were undertaken to examine modeling alternatives, sources of error, and available analysis techniques. Extensive sets of data were amassed and evaluated as part of projects to establish seismic input parameters and to produce the fragility curves. (author)

Seismic Hazard Assessment for a Characteristic Earthquake Scenario: Probabilistic-Deterministic Method

Science.gov (United States)

mouloud, Hamidatou

2016-04-01

The objective of this paper is to analyze the seismic activity and the statistical treatment of seismicity catalog the Constantine region between 1357 and 2014 with 7007 seismic event. Our research is a contribution to improving the seismic risk management by evaluating the seismic hazard in the North-East Algeria. In the present study, Earthquake hazard maps for the Constantine region are calculated. Probabilistic seismic hazard analysis (PSHA) is classically performed through the Cornell approach by using a uniform earthquake distribution over the source area and a given magnitude range. This study aims at extending the PSHA approach to the case of a characteristic earthquake scenario associated with an active fault. The approach integrates PSHA with a high-frequency deterministic technique for the prediction of peak and spectral ground motion parameters in a characteristic earthquake. The method is based on the site-dependent evaluation of the probability of exceedance for the chosen strong-motion parameter. We proposed five sismotectonique zones. Four steps are necessary: (i) identification of potential sources of future earthquakes, (ii) assessment of their geological, geophysical and geometric, (iii) identification of the attenuation pattern of seismic motion, (iv) calculation of the hazard at a site and finally (v) hazard mapping for a region. In this study, the procedure of the earthquake hazard evaluation recently developed by Kijko and Sellevoll (1992) is used to estimate seismic hazard parameters in the northern part of Algeria.

Nucelar reactor seismic safety analysis techniques

International Nuclear Information System (INIS)

Cummings, G.E.; Wells, J.E.; Lewis, L.C.

1979-04-01

In order to provide insights into the seismic safety requirements for nuclear power plants, a probabilistic based systems model and computational procedure have been developed. This model and computational procedure will be used to identify where data and modeling uncertainties need to be decreased by studying the effect of these uncertainties on the probability of radioactive release and the probability of failure of various structures, systems, and components. From the estimates of failure and release probabilities and their uncertainties the most sensitive steps in the seismic methodologies can be identified. In addition, the procedure will measure the uncertainty due to random occurrences, e.g. seismic event probabilities, material property variability, etc. The paper discusses the elements of this systems model and computational procedure, the event-tree/fault-tree development, and the statistical techniques to be employed

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 ...

Delineation of seismic source zones based on seismicity parameters ...

Indian Academy of Sciences (India)

these source zones were evaluated and were used in the hazard evaluation. ... seismic sources, linear and areal, were considered in the present study to model the seismic sources in the ..... taken as an authentic reference manual for iden-.

Seismic re-evaluation criteria for Bohunice V1 reconstruction

International Nuclear Information System (INIS)

Campbell, R.; Schlund, H.; Warnken, L.

2001-01-01

Bohunice V1 in Slovakia is a Russian designed two unit WWER 440, Model 230 Pressurized Water Reactor. The plant was not originally designed for earthquake. Subsequent and ongoing reassessments now confirm that the seismic hazard at the site is significant. EBO, the plant owner has contracted with a consortium lead by Siemens AG (REKON) to do major reconstruction of the plant to significantly enhance its safety systems by the addition of new systems and the upgrading of existing systems. As part of the reconstruction, a complete seismic assessment and upgrading is required for existing safety relevant structures, systems and components. It is not practical to conduct this reassessment and upgrading using criteria applied to new design of nuclear power plants. Alternate criteria may be used to achieve adequate safety goals. Utilities in the U.S. have faced several seismic issues with operating NPPs and to resolve these issues, alternate criteria have been developed which are much more cost effective than use of criteria for new design. These alternate criteria incorporate the knowledge obtained from investigation of the performance of equipment in major earthquakes and include provisions for structures and passive equipment to deform beyond the yield point, yet still provide their essential function. IAEA has incorporated features of these alternate criteria into draft Technical Guidelines for application to Bohunice V1 and V2. REKON has developed plant specific criteria and procedures for the Bohunice V1 reconstruction that incorporate major features of the U.S. developed alternate criteria, comply to local codes and which envelop the draft IAEA Technical Guidelines. Included in these criteria and procedures are comprehensive walkdown screening criteria for equipment, piping, HVAC and cable raceways, analytical criteria which include inelastic energy absorption factors defined on an element basis and testing criteria which include specific guidance on interpretation

One-dimensional Seismic Analysis of a Solid-Waste Landfill

International Nuclear Information System (INIS)

Castelli, Francesco; Lentini, Valentina; Maugeri, Michele

2008-01-01

Analysis of the seismic performance of solid waste landfill follows generally the same procedures for the design of embankment dams, even if the methods and safety requirements should be different. The characterization of waste properties for seismic design is difficult due the heterogeneity of the material, requiring the procurement of large samples. The dynamic characteristics of solid waste materials play an important role on the seismic response of landfill, and it also is important to assess the dynamic shear strengths of liner materials due the effect of inertial forces in the refuse mass. In the paper the numerical results of a dynamic analysis are reported and analysed to determine the reliability of the common practice of using 1D analysis to evaluate the seismic response of a municipal solid-waste landfill. Numerical results indicate that the seismic response of a landfill can vary significantly due to reasonable variations of waste properties, fill heights, site conditions, and design rock motions

Seismic and tsunami safety margin assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

Seismic and tsunami safety margin assessment

International Nuclear Information System (INIS)

2013-01-01

Nuclear Regulation Authority is going to establish new seismic and tsunami safety guidelines to increase the safety of NPPs. The main purpose of this research is testing structures/components important to safety and tsunami resistant structures/components, and evaluating the capacity of them against earthquake and tsunami. Those capacity data will be utilized for the seismic and tsunami back-fit review based on the new seismic and tsunami safety guidelines. The summary of the program in 2012 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. PWR emergency diesel generator partial-model seismic capacity tests have been conducted and quantitative seismic capacities have been evaluated. 2. Seismic capacity evaluation of switching-station electric equipment. Existing seismic test data investigation, specification survey and seismic response analyses have been conducted. 3. Tsunami capacity evaluation of anti-inundation measure facilities. Tsunami pressure test have been conducted utilizing a small breakwater model and evaluated basic characteristics of tsunami pressure against seawall structure. (author)

Destructiveness criteria for seismic risk evaluation of nuclear power plant

International Nuclear Information System (INIS)

Saragoni, G.R.

1995-01-01

Two criteria of destructiveness for seismic risk evaluation of nuclear power plant are presented. The first one is a simple linear criterion that allows to compute average response spectra in terms of earthquake accelerogram characteristics. The second defines the destructiveness potential factor P D which measures the capacity of earthquake to produce nonlinear damage. This second criterion that shows large differences of destructiveness capacity for earthquake accelerograms of different seismic environment, specially between subductive and transcursive, is strongly recommended. (author). 8 refs., 1 fig. 1 tab

Seismic evaluation of Tank 241C106 in support of retrieval activities

International Nuclear Information System (INIS)

Wallace, D.A.

1994-01-01

Tank 241C106 (C106) is a domed, single-shell high-level waste storage tank that has been in service in the 200 East Area of the Hanford Site since 1947. Tank C106 is one of twelve tanks in a 4 x 3 array with a 100-ft center-to-center spacing. Each of the tanks is approximately 75 ft in diameter, 24-ft high at the haunch, and 33-ft high at the dome apex. The level of waste in C106 and the associated thermal environment have varied throughout the life of the tanks with the peak temperature in the concrete reaching approximately 300 F at the base of the tank in the mid-1970's (Bander 1992). The calculated peak temperature in the concrete has decreased since that time to approximately 200 F. The peak temperature occurs at the inside bottom of the tank; concrete temperatures in the wall and dome are less than 130 F. The waste inside the tank is primarily solid matter approximately 7- to 8-ft deep. The tank is completely buried in dry, sandy soil to a depth of approximately 6 ft at the dome apex. The in situ evaluation of C106 documented in July 1994 includes only the effects of gravity and thermal loads. A preliminary seismic evaluation of C106 considering only horizontal excitation demonstrated the finite-element program SASSI (A System for Analysis of Soil-Structure Interaction) and provided an estimate of seismic effects including soil-to-structure interaction. This final seismic evaluation expands on the preliminary seismic evaluation to include further verification and refinement of analysis parameters, quantification to tank-to-tank and waste-to-tank interaction, and examination of the effects of vertical seismic excitation. The concrete structure of tank C106 is classified as a Safety Class 1 non-reactor structure

Wind/seismic comparison for upgrading existing structures

International Nuclear Information System (INIS)

Giller, R.A.

1989-01-01

This paper depicts the analysis procedures and methods used to evaluate three existing building structures for extreme wind loads. The three structures involved in this evaluation are located at the US Department of Energy's Hanford Site near Richland, Washington. This site is characterized by open flat grassland with few surrounding obstructions and has extreme winds in lieu of tornados as a design basis accident condition. This group of buildings represents a variety of construction types, including a concrete stack, a concrete load-bearing wall structure, and a rigid steel-frame building. The three structures included in this group have recently been evaluated for response to the design basis earthquake that included non-linear time history effects. The resulting loads and stresses from the wind analyses were compared to the loads and stresses resulting from seismic analyses. This approach eliminated the need to prepare additional capacity calculations that were already contained in the seismic evaluations

Seismic Performance Evaluation of Steel Moment Resisting Frames with WUF-B Connections

International Nuclear Information System (INIS)

Moon, Ki-Hoon; Han, Sang-Whan

2008-01-01

The purpose of this study is to evaluate the seismic performance of the moment resisting steel frames having Welded Unreinforced Flange-Bolted web (WUF-B) connections. The connections are designed and detailed in compliance with FEMA 350 recommended seismic design criteria. To conduct the seismic performance evaluation this study developed an analytical model for the pre- and post-Northridge connections based on test results. Three different frames are considered which have three-, nine- and twenty-story. Incremental Dynamic Analysis (IDA is conducted to estimate limit state capacities The performance of the frames having either pre- or post-Northridge connections is compared with the corresponding frame with ductile connections which do not experience connection fracture. The analytical results showed that buildings with post-Northridge WUF-B connections provide superior strength and interstory drift ratio capacity than buildings with pre-Northridge WUF-B connections

Seismic evaluation of a hot cell structure

International Nuclear Information System (INIS)

Srinivasan, M.G.; Kot, C.A.

1995-01-01

The evaluation of the structural capacity of and the seismic demand on an existing hot cell structure in a nuclear facility is described. An ANSYS finite-element model of the cell was constructed, treating the walls as plates and the floor and ceiling as a system of discrete beams. A modal analysis showed that the fundamental frequencies of the cell walls lie far above the earthquake frequency range. An equivalent static analysis of the structure was performed. Based on the analysis it was demonstrated that the hot cell structure, would readily withstand the evaluation basis earthquake

Bayesian methodology for generic seismic fragility evaluation of components in nuclear power plants

International Nuclear Information System (INIS)

Yamaguchi, Akira; Campbell, R.D.; Ravindra, M.K.

1991-01-01

Bayesian methodology for updating the seismic fragility of components in nuclear power plants is presented. The generic fragility data which have been evaluated based on the past SPSAs are combined with the seismic experience data. Although the seismic experience is limited to the acceleration range below the median capacity of the components, it has been found that the evidence is effective to update the fragility tail. In other words, the uncertainty of the fragility is reduced although the median capacity itself is not modified to a great extent. The annual frequency of failure is also reduced as a result of the updating of the fragility tail. The PDF of the seismic capacity is handled in discrete form, which enables the use of arbitrary type of prior distribution. Accordingly, the Log-N prior can be used which is consistent with the widely used fragility model. For evaluating posterior fragility parameters (A m and B U ), two methods have been proposed. Furthermore, it has been found that the importance of evidence used in the Bayesian methodology can be quantified by the entropy of the evidence. Only the events with high entropy need to be considered in the Bayesian updating of the fragility. The currently available seismic experience database for typical components can be utilized to develop the fragility tail which is contributive to the seismically-induced failure frequency. The combined use of generic fragility and seismic experience data, with the aid of Bayesian methodology, provides refined generic fragility curves which are useful for SPSA studies. (author)

Seismic damage assessment of reinforced concrete containment structures

International Nuclear Information System (INIS)

Cho, HoHyun; Koh, Hyun-Moo; Hyun, Chang-Hun; Kim, Moon-Soo; Shin, Hyun Mock

2003-01-01

This paper presents a procedure for assessing seismic damage of concrete containment structures using the nonlinear time-history numerical analysis. For this purpose, two kinds of damage index are introduced at finite element and structural levels. Nonlinear finite element analysis for the containment structure applies PSC shell elements using a layered approach leading to damage indices at finite element and structural levels, which are then used to assess the seismic damage of the containment structure. As an example of such seismic damage assessment, seismic damages of the containment structure of Wolsong I nuclear power plant in Korea are evaluated against 30 artificial earthquakes generated with a wide range of PGA according to US NRC regulatory guide 1.60. Structural responses and corresponding damage index according to the level of PGA and nonlinearity are investigated. It is also shown that the containment structure behaves elastically for earthquakes corresponding to or lower than DBE. (author)

Ground motion input in seismic evaluation studies

International Nuclear Information System (INIS)

Sewell, R.T.; Wu, S.C.

1996-07-01

This report documents research pertaining to conservatism and variability in seismic risk estimates. Specifically, it examines whether or not artificial motions produce unrealistic evaluation demands, i.e., demands significantly inconsistent with those expected from real earthquake motions. To study these issues, two types of artificial motions are considered: (a) motions with smooth response spectra, and (b) motions with realistic variations in spectral amplitude across vibration frequency. For both types of artificial motion, time histories are generated to match target spectral shapes. For comparison, empirical motions representative of those that might result from strong earthquakes in the Eastern U.S. are also considered. The study findings suggest that artificial motions resulting from typical simulation approaches (aimed at matching a given target spectrum) are generally adequate and appropriate in representing the peak-response demands that may be induced in linear structures and equipment responding to real earthquake motions. Also, given similar input Fourier energies at high-frequencies, levels of input Fourier energy at low frequencies observed for artificial motions are substantially similar to those levels noted in real earthquake motions. In addition, the study reveals specific problems resulting from the application of Western U.S. type motions for seismic evaluation of Eastern U.S. nuclear power plants

Generalized Fragility Relationships with Local Site Conditions for Probabilistic Performance-based Seismic Risk Assessment of Bridge Inventories

Directory of Open Access Journals (Sweden)

Sivathayalan S.

2012-01-01

Full Text Available The current practice of detailed seismic risk assessment cannot be easily applied to all the bridges in a large transportation networks due to limited resources. This paper presents a new approach for seismic risk assessment of large bridge inventories in a city or national bridge network based on the framework of probabilistic performance based seismic risk assessment. To account for the influences of local site effects, a procedure to generate site-specific hazard curves that includes seismic hazard microzonation information has been developed for seismic risk assessment of bridge inventories. Simulated ground motions compatible with the site specific seismic hazard are used as input excitations in nonlinear time history analysis of representative bridges for calibration. A normalizing procedure to obtain generalized fragility relationships in terms of structural characteristic parameters of bridge span and size and longitudinal and transverse reinforcement ratios is presented. The seismic risk of bridges in a large inventory can then be easily evaluated using the normalized fragility relationships without the requirement of carrying out detailed nonlinear time history analysis.

Evaluation of seismic margins for an in-plant piping system

International Nuclear Information System (INIS)

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

1991-01-01

Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

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

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

A scenario-based procedure for seismic risk analysis

International Nuclear Information System (INIS)

Kluegel, J.-U.; Mualchin, L.; Panza, G.F.

2006-12-01

A new methodology for seismic risk analysis based on probabilistic interpretation of deterministic or scenario-based hazard analysis, in full compliance with the likelihood principle and therefore meeting the requirements of modern risk analysis, has been developed. The proposed methodology can easily be adjusted to deliver its output in a format required for safety analysts and civil engineers. The scenario-based approach allows the incorporation of all available information collected in a geological, seismotectonic and geotechnical database of the site of interest as well as advanced physical modelling techniques to provide a reliable and robust deterministic design basis for civil infrastructures. The robustness of this approach is of special importance for critical infrastructures. At the same time a scenario-based seismic hazard analysis allows the development of the required input for probabilistic risk assessment (PRA) as required by safety analysts and insurance companies. The scenario-based approach removes the ambiguity in the results of probabilistic seismic hazard analysis (PSHA) which relies on the projections of Gutenberg-Richter (G-R) equation. The problems in the validity of G-R projections, because of incomplete to total absence of data for making the projections, are still unresolved. Consequently, the information from G-R must not be used in decisions for design of critical structures or critical elements in a structure. The scenario-based methodology is strictly based on observable facts and data and complemented by physical modelling techniques, which can be submitted to a formalised validation process. By means of sensitivity analysis, knowledge gaps related to lack of data can be dealt with easily, due to the limited amount of scenarios to be investigated. The proposed seismic risk analysis can be used with confidence for planning, insurance and engineering applications. (author)

Seismic evaluation and strengthening of Bohunice nuclear power plant structures

International Nuclear Information System (INIS)

Shipp, J.G.; Short, S.A.; Grief, T.; Borov, V.; Kuzma, J.

2001-01-01

A seismic assessment and strengthening investigation is being performed for selected structures at the Bohunice V1 Nuclear Power Plant in Slovakia. Structures covered in this paper include the reactor building complex and the emergency generator station. The emergency generator station is emphasized in the paper as work is nearly complete while work on the reactor building complex is ongoing at this time. Seismic evaluation and strengthening work is being performed by a cooperative effort of Siemens and EQE along with local contractors. Seismic input is the interim Review Level Earthquake (horizontal peak ground acceleration of 0.3 g). The Bohunice V1 reactor building complex is a WWER 4401230 nuclear power plant that was originally built in the mid-1970s but had extensive seismic upgrades in 1991. Siemens has performed three dimensional dynamic analyses of the reactor building complex to develop seismic demand in structural elements. EQE is assessing seismic capacities of structural elements and developing strengthening schemes, where needed. Based on recent seismic response analyses for the interim Review Level Earthquake which account for soil-structure interaction in a rigorous manner, the 1991 seismic upgrade has been found to be inadequate in both member/connection strength and in providing complete load paths to the foundation. Additional strengthening is being developed. The emergency generator station was built in the 1970s and is a two-story unreinforced brick masonry (URM) shear wall building above grade with a one story reinforced concrete shear wall basement below grade. Seismic analyses and testing of the URM walls has been performed to assess the need for building strengthening. Required structural strengthening for in-plane forces consists of revised and additional vertical steel framing and connections, stiffening of horizontal roof bracing, and steel connections between the roof and supporting walls and pointing of two interior transverse URM

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)

A seismic study on cracks in crystalline rock

International Nuclear Information System (INIS)

Israelsson, H.

1981-07-01

This report summarizes results from a field study with in-situ seismic measurements in crystalline rock. It was found that among a few potential seismic techniques the so called cross hole method would probably provide the most powerful capability for detecting cracks and fracture zones. By this method the area between two holes are systematically scanned by seismic raypaths. Seismic signals are generated in one hole by micro explosions and recorded in the other at various combinations of depths. A test sample of scanning data showed a rather dramatic variation of the seismic P-wave velocity (5-6 km/s). Analysis procedures like tomographic imaging was applied to this data set primarily to illustrate the kind of structural mapping such procedures can provide. (Author)

An experimental scale-model study of seismic response of an underground opening in jointed rock mass

International Nuclear Information System (INIS)

Kana, D.D.; Fox, D.J.; Hsiung, S.; Chowdhury, A.H.

1997-02-01

This report describes an experimental investigation conducted by the Center for Nuclear Waste Regulatory Analyses (CNWRA) to (i) obtain a better understanding of the seismic response of an underground opening in a highly-fractured and jointed rock mass and (ii) generate a data set that can be used to evaluate the capabilities (analytical methods) to calculate such response. This report describes the design and implementation of simulated seismic experiments and results for a 1/15 scale model of a jointed rock mass with a circular tunnel in the middle. The discussion on the design of the scale model includes a description of the associated similitude theory, physical design rationale, model material development, preliminary analytical evaluation, instrumentation design and calibration, and model assembly and pretest procedures. The thrust of this discussion is intended to provide the information necessary to understand the experimental setup and to provide the background necessary to understand the experimental results. The discussion on the experimental procedures and results includes the seismic input test procedures, test runs, and measured excitation and response time histories. The closure of the tunnel due to various levels of seismic activity is presented. A threshold level of seismic input amplitude was required before significant rock mass motion occurred. The experiment, though designed as a two-dimensional representation of a rock mass, behaved in a somewhat three-dimensional manner, which will have an effect on subsequent analytical model comparison

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)

Reprocessing and interpretation, seismic reflection data: Hanford Site, Pasco Basin, south central Washington

International Nuclear Information System (INIS)

Berkman, E.

1983-01-01

The purpose of this project was to reprocess, evaluate, and reinterpret 14 line miles of seismic reflection data acquired at the Hanford Site. Regional and area-specific geology has been reviewed, the data acquisition parameters as they relate to the limitations inherent in the data have been discussed, and the reprocessing procedures have been described in detail along with an evaluation of the original processing. After initial testing, the focus of the reprocessing was placed on resolution of the geologic horizons at and near the top of the basalt. The reprocessed seismic data shows significant improvement over the original processing. The improvement is the result of the integrated processing and interpretation approach where each processing step has been tested in sequence and the intermediate results examined carefully in accordance with the project goals. The interpretation procedure placed strong reliance upon synthetic seismograms and models calculated based upon the physical parameters of the subsurface materials, and upon associated geophysical (reflection, gravity, magnetic) data. The final interpretation of the seismic data is in agreement with the structural contour maps based primarily on borehole information. The seismic interpretation has added important detail concerning areas which should be considered for further study. 60 figs., 1 tab

Performance-based seismic assessment of vulnerability of dam using time history analysis

Directory of Open Access Journals (Sweden)

Elmrabet Oumnia

2018-01-01

Full Text Available The current performance-based seismic assessment procedure can be computationally intensive as it requires many time history analyses (THA each requiring time intensive post-processing of results. Time history analysis is a part of structural analysis and is the calculation of the response of a structure to any earthquake. It is one of the main processes of structural design in regions where earthquakes are prevalent. The objective of this study is to evaluate the seismic performance of embankment dam located on the Oued RHISS in the Province of AL HOCEIMA using the THA method. To monitor structural behavior, the seismic vulnerability of structure is evaluated under real earthquake records with considering the soil-structure-fluide interaction. In this study, a simple assistant program is developed for implementing earthquake analyses of structure with ANSYS, ground acceleration–time history data are used for seismic analysis and dynamic numerical simulations were conducted to study and identify the total response of the soil-structure system.

A regulatory view of the seismic re-evaluation of existing nuclear power plants in the United Kingdom

International Nuclear Information System (INIS)

Inkester, J.E.; Bradford, P.M.

1995-01-01

The paper describes the background to the seismic re-evaluation of existing nuclear power plants in the United Kingdom. Nuclear installations in this country were not designed specifically to resist earthquakes until the nineteen-seventies, although older plants were robustly constructed. The seismic capability of these older installations is now being evaluated as part of the periodic safety reviews which nuclear licensees are required to carry out. The regulatory requirements which set the framework for these studies are explained and the approaches being adopted by the licensees for their assessment of the seismic capability of existing plants are outlined. The process of hazard appraisal is reported together with a general overview of UK seismicity. The paper then discusses the methodologies used to evaluate the response of plant to the hazard. Various other types of nuclear installation besides power plants are subject to licensing in the UK and the application of seismic evaluation to some of these is briefly described. Finally the paper provides some comments on future initiatives and possible areas of development. (author)

Seismic evaluation of existing nuclear power plants and other facilities V. 2. Proceedings of the technical committee meeting. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-01-01

The objectives of this TCM are: to review the IAEA Safety Report on Seismic Evaluation of Existing Nuclear Power Plants in order to achieve a consensus among Member States on this matter and to discuss the outlines of an IAEA Co-ordinated Research Programme on specific topics related to this subject. This volume includes presentations of the member states describing the practical approach to evaluation of seismic equipment of the existing NPPs, validation of innovative systems for earthquake protection; seismic re-evaluation of the NPPs, seismic regulations and safety standards; and other activities related to seismic safety in Member States.

Seismic evaluation of existing nuclear power plants and other facilities V. 2. Proceedings of the technical committee meeting. Working material

International Nuclear Information System (INIS)

2002-01-01

The objectives of this TCM are: to review the IAEA Safety Report on Seismic Evaluation of Existing Nuclear Power Plants in order to achieve a consensus among Member States on this matter and to discuss the outlines of an IAEA Co-ordinated Research Programme on specific topics related to this subject. This volume includes presentations of the member states describing the practical approach to evaluation of seismic equipment of the existing NPPs, validation of innovative systems for earthquake protection; seismic re-evaluation of the NPPs, seismic regulations and safety standards; and other activities related to seismic safety in Member States

Task force activity to take the effect of elastic-plastic behaviour into account on the seismic safety evaluation of nuclear piping systems

International Nuclear Information System (INIS)

Nakamura, Izumi; Shiratori, Masaki; Morishita, Masaki; Otani, Akihito; Shibutani, Tadahito

2015-01-01

According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure. Since the stress assessment based on the elastic analysis does not reflect actual seismic capability of piping systems including plastic region, it is necessary to develop a rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load. With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a task force activity has been planned. Through the activity, the authors intend to establish guidelines to estimate the elastic-plastic behavior of piping systems rationally and conservatively, and to provide new rational seismic safety criteria taking the effect of elastic-plastic behavior into account. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test. In this paper, the outline of the research activity and the preliminary results of benchmark analyses are described. (author)

Seismic fatigue life evaluation of mechanical structures using energy balance equation

International Nuclear Information System (INIS)

Minagawa, Keisuke; Fujita, Satoshi; Kitamura, Seiji; Okamura, Shigeki

2009-01-01

Evaluation of seismic resistant performance for severe earthquakes is required, because of occurrence of earthquakes which exceed the design criteria. Additionally, quantitative evaluation of cumulative damage by earthquake is also required. In this study, the energy balance equation is applied to the evaluation. The energy balance equation expresses integral information of response, so that the energy balance equation is adequate for the evaluation of the influence of cumulative load such as seismic response. At first, vibration experiment that leads experimental model to fatigue failure by continuous vibration disturbance is conducted. As a result of the experiment, relation between fatigue failure and energy balance equation is confirmed. Then the relation is proved from the viewpoint of hysteresis energy, and consistency between energy balance equation and hysteresis energy is confirmed. Finally, we adopted cumulative damage rule to energy balance equation in order to expect the fatigue life under random waves that have various input acceleration. (author)

Seismic Evaluation of Hydrocarbon Saturation in Deep-Water Reservoirs

Energy Technology Data Exchange (ETDEWEB)

Michael Batzle

2006-04-30

During this last period of the ''Seismic Evaluation of Hydrocarbon Saturation in Deep-Water Reservoirs'' project (Grant/Cooperative Agreement DE-FC26-02NT15342), we finalized integration of rock physics, well log analysis, seismic processing, and forward modeling techniques. Most of the last quarter was spent combining the results from the principal investigators and come to some final conclusions about the project. Also much of the effort was directed towards technology transfer through the Direct Hydrocarbon Indicators mini-symposium at UH and through publications. As a result we have: (1) Tested a new method to directly invert reservoir properties, water saturation, Sw, and porosity from seismic AVO attributes; (2) Constrained the seismic response based on fluid and rock property correlations; (3) Reprocessed seismic data from Ursa field; (4) Compared thin layer property distributions and averaging on AVO response; (5) Related pressures and sorting effects on porosity and their influence on DHI's; (6) Examined and compared gas saturation effects for deep and shallow reservoirs; (7) Performed forward modeling using geobodies from deepwater outcrops; (8) Documented velocities for deepwater sediments; (9) Continued incorporating outcrop descriptive models in seismic forward models; (10) Held an open DHI symposium to present the final results of the project; (11) Relations between Sw, porosity, and AVO attributes; (12) Models of Complex, Layered Reservoirs; and (14) Technology transfer Several factors can contribute to limit our ability to extract accurate hydrocarbon saturations in deep water environments. Rock and fluid properties are one factor, since, for example, hydrocarbon properties will be considerably different with great depths (high pressure) when compared to shallow properties. Significant over pressure, on the other hand will make the rocks behave as if they were shallower. In addition to the physical properties, the scale and

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.

Study on structural seismic margin and probabilistic seismic risk. Development of a structural capacity-seismic risk diagram

International Nuclear Information System (INIS)

Nakajima, Masato; Ohtori, Yasuki; Hirata, Kazuta

2010-01-01

Seismic margin is extremely important index and information when we evaluate and account seismic safety of critical structures, systems and components quantitatively. Therefore, it is required that electric power companies evaluate the seismic margin of each plant in back-check of nuclear power plants in Japan. The seismic margin of structures is usually defined as a structural capacity margin corresponding to design earthquake ground motion. However, there is little agreement as to the definition of the seismic margin and we have no knowledge about a relationship between the seismic margin and seismic risk (annual failure probability) which is obtained in PSA (Probabilistic Safety Assessment). The purpose of this report is to discuss a definition of structural seismic margin and to develop a diagram which can identify a relation between seismic margin and seismic risk. The main results of this paper are described as follows: (1) We develop seismic margin which is defined based on the fact that intensity of earthquake ground motion is more appropriate than the conventional definition (i.e., the response-based seismic margin) for the following reasons: -seismic margin based on earthquake ground motion is invariant where different typed structures are considered, -stakeholders can understand the seismic margin based on the earthquake ground motion better than the response-based one. (2) The developed seismic margin-risk diagram facilitates us to judge easily whether we need to perform detailed probabilistic risk analysis or only deterministic analysis, given that the reference risk level although information on the uncertainty parameter beta is not obtained. (3) We have performed numerical simulations based on the developed method for four sites in Japan. The structural capacity-risk diagram differs depending on each location because the diagram is greatly influenced by seismic hazard information for a target site. Furthermore, the required structural capacity

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.

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

Structural evaluation of the 2736Z Building for seismic loads

International Nuclear Information System (INIS)

Giller, R.A.

1994-01-01

The 2736Z building structure is evaluated for high-hazard loads. The 2736Z building is analyzed herein for normal and seismic loads and is found to successfully meet the guidelines of UCRL-15910 along with the related codes requirements

Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

International Nuclear Information System (INIS)

Arthur, D.F.; Dong, R.G.; Murray, R.C.; Nelson, T.A.; Smith, P.D.; Wight, L.H.

1978-01-01

Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

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.

Evaluation of granular soil properties in seismic analysis of nuclear structures

International Nuclear Information System (INIS)

Bica, A.; Riera, J.D.; Nanni, L.F.

1983-01-01

The seismic analysis of nuclear power plant structures founded on soils, as well as related soil-structure interaction studies, are often made by means of 'equivalent' linear models of soil behavior, represented by effective values of damping and of Young's modulus. Such approach requires resorting to iteration on the material properties, thus leading to a 'multilinear' analysis which can be justified in practice on account of the scarce knowledge of constitutive equations applicable to soils under a general three-dimensional stress state. It is therefore important to establish bounds on the applicability of the multilinear solutions, and to develop reliable procedures for the evaluation of the soil properties to be used in seismic analyses. The paper focuses attention on the dynamic properties of sandy soils. To that effect, an extensive program was conducted using a triaxial dynamic testing apparatus developed at the UFRGS, and the results compared with existing experimental evidence, including data from resonant-column testing. Linear and nonlinear regression techniques applied to the experimental data led to new equations relating damping and soil stiffness to the dependent variables, and permitted as well the determination of the expected error of the estimated parameters. It was found that an increasing frequency, slightly increases both Young's modulus and the effective damping ratio. In addition, the influence of the content of fines was found to be significant. This variable does not appear in several available empirical equations, which only consider the confining pressure, the void ratio and the amplitude of the cyclic shear deformations as relevant variables. (orig.)

Enhancement and feature extraction of RS images from seismic area and seismic disaster recognition technologies

Science.gov (United States)

Zhang, Jingfa; Qin, Qiming

2003-09-01

Many types of feature extracting of RS image are analyzed, and the work procedure of pattern recognizing in RS images of seismic disaster is proposed. The aerial RS image of Tangshan Great Earthquake is processed, and the digital features of various typical seismic disaster on the RS image is calculated.

Source modelling in seismic risk analysis for nuclear power plants

International Nuclear Information System (INIS)

Yucemen, M.S.

1978-12-01

The proposed probabilistic procedure provides a consistent method for the modelling, analysis and updating of uncertainties that are involved in the seismic risk analysis for nuclear power plants. The potential earthquake activity zones are idealized as point, line or area sources. For these seismic source types, expressions to evaluate their contribution to seismic risk are derived, considering all the possible site-source configurations. The seismic risk at a site is found to depend not only on the inherent randomness of the earthquake occurrences with respect to magnitude, time and space, but also on the uncertainties associated with the predicted values of the seismic and geometric parameters, as well as the uncertainty in the attenuation model. The uncertainty due to the attenuation equation is incorporated into the analysis through the use of random correction factors. The influence of the uncertainty resulting from the insufficient information on the seismic parameters and source geometry is introduced into the analysis by computing a mean risk curve averaged over the various alternative assumptions on the parameters and source geometry. Seismic risk analysis is carried for the city of Denizli, which is located in the seismically most active zone of Turkey. The second analysis is for Akkuyu

Non-linear transient behavior during soil liquefaction based on re-evaluation of seismic records

OpenAIRE

Kamagata, S.; Takewaki, Izuru

2015-01-01

Focusing on soil liquefaction, the seismic records during the Niigata-ken earthquake in 1964, the southern Hyogo prefecture earthquake in 1995 and the 2011 off the Pacific coast of Tohoku earthquake are analyzed by the non-stationary Fourier spectra. The shift of dominant frequency in the seismic record of Kawagishi-cho during the Niigata-ken earthquake is evaluated based on the time-variant property of dominant frequencies. The reduction ratio of the soil stiffness is evaluated from the shif...

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

Korea-Japan Joint Research on Development of Seismic Capacity Evaluation and Enhancement Technology Considering Near-Fault Effect

Energy Technology Data Exchange (ETDEWEB)

Choun, Young Sun; Choi, In Kil; Kim, Min Kyu [KAERI, Daejon (Korea, Republic of); Ohtori, Yasuki; Shiba, Yoshiaki; Nakajima, Masato [Central Research Institute of Electric Power Industry, Tokyo (Japan)

2005-12-15

Several recent improved methods for the EGFM are introduced in order to avoid artificial holes seen in the synthetic acceleration spectrum. Furthermore evaluation of input ground motions at Wolsung NPP are performed by varying the source parameters that may control the high-frequency wave radiation and the deviation of the synthetic motions are revealed. The PSHA case studies for four NPP sites (Wolsung, Kori, Uljin, Younggwang) are performed. In the analysis, site-specific attenuation equations developed for Korean NPP sites are employed, and the seismic hazards for the target sites are evaluated in the case where the four kind of seismic source models are considered. Moreover, the PSHA for Wolsung and Younggwang are conducted by using the site-specific attenuation equation with the index of response spectra and the uniform hazard spectra are evaluated for the two sites. The supporting tool for seismic response analysis and the evaluation tool for evaluating annual probability of failure were integrated in the frame of the seismic risk assessment system. Then, the tools were applied to the seismic risk assessment of the conventional EDG and isolated EDG. General information such as earthquake parameters and regional distribution of seismic intensity is summarized on the 2005 West Off Fukuoka earthquake. Then, the observed strong motion records in Japan and Korea sites are compiled, and regional distribution of peak accelerations are represented. Moreover, the peak accelerations of the records are compared with the values estimated from the existing attenuation equations.

Equivalent viscous damping procedure for multi-material systems

International Nuclear Information System (INIS)

Ahmed, H.; Ma, D.

1979-01-01

The inclusion of accurate viscous damping effects in the seismic analysis of nuclear power plants is discussed. A procedure to evaluate and use equivalent viscous damping coefficients in conjunction with the substructure method of finite element analysis is outlined in detail

Mechanical property test of natural rubber bearing for the evaluation of uncertainty value of seismic isolation devices

International Nuclear Information System (INIS)

Kim, Min Kyu; Kim, Jung Han; Choi, In Kil

2012-01-01

Seismic safety of NPP is one of the most important issues in a nuclear field after great east Japan earthquake in 2011. For the improvement of seismic safety of nuclear power plant, seismic isolation is the easiest solution for increasing the seismic safety. Otherwise, the application of seismic isolation devices for nuclear power plants doesn't make the seismic risk of NPP increases always. The rubber bearing have many uncertainties of material properties and large displacement should absorb according to the application of isolation devices. In this study, for the evaluation of uncertainty of the material properties of rubber bearing, material tests for rubber and mechanical properties test for natural rubber bearing were performed. For the evaluation of effect of hardness of rubber, 4 kinds of rubber hardness for material property tests and 2 kinds of rubber hardness for mechanical property test were considered. As a result, the variation of material properties is higher than that of mechanical properties of natural rubber bearings

Evaluating Seismic Site Effects at Cultural Heritage Sites in the Mediterranean Area

Science.gov (United States)

Imposa, S.; D'Amico, S.; Panzera, F.; Lombardo, G.; Grassi, S.; Betti, M.; Muscat, R.

2017-12-01

Present study concern integrated geophysical and numerical simulation aiming at evaluate the seismic vulnerability of cultural heritage sites. Non-invasive analysis targeted to characterize local site effects as well as dynamic properties of the structure were performed. Data were collected at several locations in the Maltese Archipelago (central Mediterranean) and in some historical buildings located in Catania (Sicily). In particular, passive seismic techniques and H/V data where used to derive 1D velocity models and amplification functions. The dynamic properties of a building are usually described through its natural frequency and the damping ratio. This latter is important in seismic design since it allows one to evaluate the ability of a structure to dissipate the vibration energy during an earthquake. The fundamental frequency of the investigated structure was obtained using ambient vibrations recorded by two or more sensors monitoring the motion at different locations in the building. Accordingly, the fundamental period of several Maltese Watchtowers and some historical buildings of Catania were obtained by computing the ratio between the amplitudes of the Fourier spectrum of horizontal (longitudinal and transverse) components recorded on the top and on the ground floors. Using ANSYS code, the modal analysis was performed to evaluate the first 50 vibration modes with the aim to check the activation of the modal masses and to assess the seismic vulnerability of the tower. The STRATA code was instead adopted in the Catania heritage buildings using as reference earthquake moderate to strong shocks that struck south-eastern Sicily. In most of the investigated buildings is was not possible to identify a single natural frequency but several oscillation modes. These results appear linked to the structural complexity of the edifices, their irregular plan shape and the presence of adjacent structures. The H/V outside the buildings were used to determine predominant

Advances in seismic criteria to qualify structures, systems and components in operating reactors

International Nuclear Information System (INIS)

Manrique, M.A.; Bak, W.R.

1989-01-01

This paper describes improved seismic evaluation criteria and analysis methodologies used as part of the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1. The plant had originally been designed for 0.25 g ground acceleration and was required to be upgraded to a 0.67 g ground acceleration as part of the plant's Long Term Service Seismic Reevaluation Program. The application of the criteria and methods described in this paper to demonstrate the seismic capability of the plant resulted in efficient plant modifications with considerable cost savings to the plant owner. The NRC accepted these criteria and methods based on favorable results of reviews, audits and independent verification of the theories, bases and implementation procedures of the proposed criteria and analysis methods

An alternative procedure for direct generation of seismic floor spectra

International Nuclear Information System (INIS)

Singh, M.P.; Sharma, A.M.

1983-01-01

Several approaches have been developed for generation of floor spectra which do not require seismic inputs in the form of acceleration time histories but can use a prescribed set of ground response spectra directly. These approaches are often referred to as direct approaches. Most of these approaches are based on the method of mode displacement of structural dynamics. Some problems can, however, occur with these approaches if the response of a system is affected by the high frequency modes and if these modes are omitted from the analyses. Herein an alternative approach based on the method of mode acceleration is proposed wherein the effect of high frequency modes is correctly included without their explicit evaluation. The seismic inputs in this approach are required to be prescribed in terms of relative acceleration and velocity spectra. The approach is very effective for the calculation of floor spectra for structural systems which have significant contribution from high frequency modes, and also for floors close to the base which are usually affected by the higher modes. In other cases too, this approach has been shown to provide better results than the mode displacement approach for a given number of modes used in the analysis. Thus, a general use of this approach in lieu of the mode displacement approach is advocated. (orig.)

Seismic Hazard Assessment in Site Evaluation for Nuclear Installations: Ground Motion Prediction Equations and Site Response

International Nuclear Information System (INIS)

2016-07-01

The objective of this publication is to provide the state-of-the-art practice and detailed technical elements related to ground motion evaluation by ground motion prediction equations (GMPEs) and site response in the context of seismic hazard assessments as recommended in IAEA Safety Standards Series No. SSG-9, Seismic Hazards in Site Evaluation for Nuclear Installations. The publication includes the basics of GMPEs, ground motion simulation, selection and adjustment of GMPEs, site characterization, and modelling of site response in order to improve seismic hazard assessment. The text aims at delineating the most important aspects of these topics (including current practices, criticalities and open problems) within a coherent framework. In particular, attention has been devoted to filling conceptual gaps. It is written as a reference text for trained users who are responsible for planning preparatory seismic hazard analyses for siting of all nuclear installations and/or providing constraints for anti-seismic design and retrofitting of existing structures

France's seismic zoning

International Nuclear Information System (INIS)

Mohammadioun, B.

1997-01-01

In order to assess the seismic hazard in France in relation to nuclear plant siting, the CEA, EDF and the BRGM (Mine and Geology Bureau) have carried out a collaboration which resulted in a seismic-tectonic map of France and a data base on seismic history (SIRENE). These studies were completed with a seismic-tectonic zoning, taking into account a very long period of time, that enabled a probabilistic evaluation of the seismic hazard in France, and that may be related to adjacent country hazard maps

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

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.

Seismic damage assessment for high-rise buildings

Science.gov (United States)

Scholl, Roger E.

1980-01-01

The problem considered in this project, conducted by URS/John A. Blume & Associates, Engineers (URS/Blume), for the U.S. Geological Survey, is the identification, evaluation, and correlation of ground-motion and structural parameters in order to improve procedures for predicting dollar losses for high-rise structures damaged by earthquakes. Ground-motion data bases, analytical techniques, and known motion-damage relationships already developed for high-rise buildings and for other classes of structures will be refined and extended so that reliable quantitative seismic risk evaluations can be made.

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)

Redistribution Principle Approach for Evaluation of Seismic Active Earth Pressure Behind Retaining Wall

Science.gov (United States)

Maskar, A. D.; Madhekar, S. N.; Phatak, D. R.

2017-11-01

The knowledge of seismic active earth pressure behind the rigid retaining wall is very essential in the design of retaining wall in earthquake prone regions. Commonly used Mononobe-Okabe (MO) method considers pseudo-static approach. Recently there are many pseudo-dynamic methods used to evaluate the seismic earth pressure. However, available pseudo-static and pseudo-dynamic methods do not incorporate the effect of wall movement on the earth pressure distribution. Dubrova (Interaction between soils and structures, Rechnoi Transport, Moscow, 1963) was the first, who considered such effect and till date, it is used for cohesionless soil, without considering the effect of seismicity. In this paper, Dubrova's model based on redistribution principle, considering the seismic effect has been developed. It is further used to compute the distribution of seismic active earth pressure, in a more realistic manner, by considering the effect of wall movement on the earth pressure, as it is displacement based method. The effects of a wide range of parameters like soil friction angle (ϕ), wall friction angle (δ), horizontal and vertical seismic acceleration coefficients (kh and kv); on seismic active earth pressure (Kae) have been studied. Results are presented for comparison of pseudo-static and pseudo-dynamic methods, to highlight the realistic, non-linearity of seismic active earth pressure distribution. The current study results in the variation of Kae with kh in the same manner as that of MO method and Choudhury and Nimbalkar (Geotech Geol Eng 24(5):1103-1113, 2006) study. To increase in ϕ, there is a reduction in static as well as seismic earth pressure. Also, by keeping constant ϕ value, as kh increases from 0 to 0.3, earth pressure increases; whereas as δ increases, active earth pressure decreases. The seismic active earth pressure coefficient (Kae) obtained from the present study is approximately same as that obtained by previous researchers. Though seismic earth

Training for the seismic qualification utility group (SQUG) generic implementation procedure (GIP) for verification of seismic effects on nuclear power plant equipment

International Nuclear Information System (INIS)

Schaffstall, R.E.; Smith, N.P.; Baker, K.M.

1993-01-01

Resolution of USI A-46 was accomplished May, 1992 with the issuance of the SER for the GIP, Revision 2. The implementation of the resolution for the SQUG plants will be completed by 1996. Both the USNRC and the SQUG utilities have expended considerable resources to effect a common understanding that the effort has been worth the price. The Training Program developed to ensure that the efforts expended by the SQUG was indeed worth the price will be a major contributor in the success of the implementation. All of the older nuclear power plants in the United States now have an alternative standard to determine seismic resistance using experience data- a standard against which they can be evaluated and found seismically adequate to achieve and maintain a safe and reliable shutdown when subjected to strong motion earthquakes

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

User's manual of a computer code for seismic hazard evaluation for assessing the threat to a facility by fault model. SHEAT-FM

International Nuclear Information System (INIS)

Sugino, Hideharu; Onizawa, Kunio; Suzuki, Masahide

2005-09-01

To establish the reliability evaluation method for aged structural component, we developed a probabilistic seismic hazard evaluation code SHEAT-FM (Seismic Hazard Evaluation for Assessing the Threat to a facility site - Fault Model) using a seismic motion prediction method based on fault model. In order to improve the seismic hazard evaluation, this code takes the latest knowledge in the field of earthquake engineering into account. For example, the code involves a group delay time of observed records and an update process model of active fault. This report describes the user's guide of SHEAT-FM, including the outline of the seismic hazard evaluation, specification of input data, sample problem for a model site, system information and execution method. (author)

Seismic risk evaluation within the technology neutral framework

International Nuclear Information System (INIS)

Johnson, B.C.; Apostolakis, G.E.

2012-01-01

Highlights: ► We examine seismic risk within the Technology Neutral Framework (TNF). ► We find that the risk goals in the TNF to be stringent compared with current goals. ► We note that the current fleet reactors would not meet the TNF goals. ► We recommend that an initiating frequency cutoff of 10 −5 per year be use in evaluating seismic risk. - Abstract: The NRC Office of Nuclear Regulatory Research has proposed a risk-informed and performance-based licensing process that is referred to as the technology neutral framework (TNF). In the TNF, licensing basis events (LBEs), determined using probabilistic risk assessment methods, take the place of design basis accidents. These LBEs are constructed by grouping together accident sequences with similar phenomenology. All event sequences with a mean frequency greater than 10 −7 per reactor year are to be considered as part of the licensing basis. Imposing such a limit would require that earthquakes with a mean return period of ten million years be considered as part of the licensing basis. It is difficult to get seismic hazards (i.e., ground accelerations) from expert seismologists at such low frequencies. This is because it is difficult or impossible to confidently say what the seismic hazard might be at these extremely low frequencies. A linear extrapolation in log-log space of hazard curves at the Clinton site down to 10 −7 per year leads to a peak ground acceleration of about 4.5 g. A Weibull distribution is also used to fit the curve leading to a peak ground acceleration of about 2.6 g. These extrapolations demonstrate the extreme nature of rare earthquakes. Even when seismic isolation is implemented, the TNF goal is not met. The problem appears to be that there is no limit on initiating event frequency in the TNF. Demonstrating that a design meets the goals of the TNF would be nearly impossible. A frequency limit for earthquakes could be imposed at a frequency of about 10 −5 per year to focus on

Walkdown procedure: Seismic adequacy review of safety class 3 ampersand 4 commodities in 2736-Z ampersand ZB buildings at PFP facility

International Nuclear Information System (INIS)

Ocoma, E.C.

1995-01-01

Seismic evaluation of existing safety class (SC) 3 and non-SC 4 commodities at the Plutonium Finishing Plant (PFP) is integrated into an area walkdown program. Field walkdowns of potential PFP seismic deficiencies associated with structural failure and falling will be performed using the DOE SQUG/EPRI methodology. Potential proximity interactions are also addressed. Objective of the walkdown is to qualify as much of the equipment as practical and to identify candidates for further evaluation

Seismic evaluation and ranking of embankments for bridges on and over the parkways in Western Kentucky.

Science.gov (United States)

2008-06-01

This study represents one of the Seismic Evaluation of Bridges on and over the Parkways in Western Kentucky investigative series. The effort is focused on the seismic vulnerability of bridge embankments against slope instability and liquefaction pote...

Procedures for evaluation of vibratory ground motions of soil deposits at nuclear power plant sites

International Nuclear Information System (INIS)

1975-06-01

According to USNRC requirements set forth in Appendix A, 10 CFR, Part 100, vibratory ground motion criteria for a nuclear plant must be based on local soil conditions, as well as on the seismicity, geology, and tectonics of the region. This report describes how such criteria can be developed by applying the latest technology associated with analytical predictions of site-dependent ground motions and with the use of composite spectra obtained from the current library of strong motion records. Recommended procedures for defining vibratory ground motion criteria contain the following steps: (1) geologic and seismologic studies; (2) site soils investigations; (3) site response sensitivity studies; (4) evaluation of local site response characteristics; (5) selection of site-matched records; and (6) appraisal and selection of seismic input criteria. An in-depth discussion of the engineering characteristics of earthquake ground motions including parameters used to characterize earthquakes and strong motion records, geologic factors that influence ground shaking, the current strong motion data base, and case histories of the effects of past earthquake events is presented. Next, geotechnical investigations of the seismologic, geologic, and site soil conditions required to develop vibratory motion criteria are briefly summarized. The current technology for establishing vibratory ground motion criteria at nuclear plant sites, including site-independent and site-dependent procedures that use data from strong motion records and from soil response analyses is described. (auth)

Review of the SQUG type seismic program at Savannah River Site

International Nuclear Information System (INIS)

Bitner, J.L.; Lin, C.W.; Anderson, N.R.; Bezler, P.

1991-01-01

The production reactors at Savannah River were shut down in 1988 because of questions about their safety. One question is whether they can withstand earthquakes. To answer the earthquake question, the site operator (Westinghouse Savannah River Company) developed a program to evaluate the capability of the safety systems in the K, L, and P reactors to function during and after an earthquake, and to upgrade them if necessary. The seismic program for Savannah River relies heavily on the Generic Implementation Procedure (GIP) developed by the Seismic qualification Utility Group. The GIP was originally developed for application to over 65 commercial power reactors throughout the U.S. It has been thoroughly reviewed by the U.S. Nuclear Regulatory Commission. The objectives of the ISWRT (Independent Seismic Walkdown Review Team) review were to: evaluate the program and evaluate its execution. The first objective was accomplished using an in-office review of the program. The second objective was accomplished using an in-office review and in-plant walkdown of selected safety systems. The ISWRT review and walkdown are summarized in this paper

Re-assessment of seismic loads in conjunction with periodic safety review

International Nuclear Information System (INIS)

Jonczyk, Josef

2002-01-01

The objective of this paper is the fundamental consideration of a safeguard-aim-oriented approach for use in the re-assessment of seismic events with regard to the periodic safety review (PSR) of nuclear power plants (NPP). The re-assessment aspects of site-specific design earthquakes (DEQ), specially the procedure for seismic hazard analysis, will not, however, be considered in detail here. The proposed assessment concept clearly presents a general approach for safety assessments. The approach is based on a successive screening review of components that are considered sufficiently earthquake-resistant. In this respect, the principle of maximum practical application of the design documentation has been considered in the re-assessment process. On the other hand, the safeguard-aim-oriented evaluation will also be applied with regard to whether the requirements of the safety regulations are fulfilled with respect to the safety goals. The review in conjunction with PSR does not, however, attempt to perform this under all technical aspects. Moreover, it is possible to make extensive use of experimental knowledge and engineering judgement with regard to the structural capacity behaviour in case of a seismic event. Compared with design procedures, however, this proposed approach differs from the one applied in licensing procedures, in which such assessment freedom will not usually be exhausted. (author)

Numerical Procedure to Forecast the Tsunami Parameters from a Database of Pre-Simulated Seismic Unit Sources

Science.gov (United States)

Jiménez, César; Carbonel, Carlos; Rojas, Joel

2018-04-01

We have implemented a numerical procedure to forecast the parameters of a tsunami, such as the arrival time of the front of the first wave and the maximum wave height in real and virtual tidal stations along the Peruvian coast, with this purpose a database of pre-computed synthetic tsunami waveforms (or Green functions) was obtained from numerical simulation of seismic unit sources (dimension: 50 × 50 km2) for subduction zones from southern Chile to northern Mexico. A bathymetry resolution of 30 arc-sec (approximately 927 m) was used. The resulting tsunami waveform is obtained from the superposition of synthetic waveforms corresponding to several seismic unit sources contained within the tsunami source geometry. The numerical procedure was applied to the Chilean tsunami of April 1, 2014. The results show a very good correlation for stations with wave amplitude greater than 1 m, in the case of the Arica tide station an error (from the maximum height of the observed and simulated waveform) of 3.5% was obtained, for Callao station the error was 12% and the largest error was in Chimbote with 53.5%, however, due to the low amplitude of the Chimbote wave (<1 m), the overestimated error, in this case, is not important for evacuation purposes. The aim of the present research is tsunami early warning, where speed is required rather than accuracy, so the results should be taken as preliminary.

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

Evaluation of Multi Canister Overpack (MCO) Handling Machine Uplift Restraint for a Seismic Event During Repositioning Operations

International Nuclear Information System (INIS)

SWENSON, C.E.

2000-01-01

Insertion of the Multi-Canister Overpack (MCO) assemblies into the Canister Storage Building (CSB) storage tubes involves the use of the MCO Handling Machine (MHM). During MCO storage tube insertion operations, inadvertent movement of the MHM is prevented by engaging seismic restraints (''active restraints'') located adjacent to both the bridge and trolley wheels. During MHM repositioning operations, the active restraints are not engaged. When the active seismic restraints are not engaged, the only functioning seismic restraints are non-engageable (''passive'') wheel uplift restraints which function only if the wheel uplift is sufficient to close the nominal 0.5-inch gap at the uplift restraint interface. The MHM was designed and analyzed in accordance with ASME NOG-1-1995. The ALSTHOM seismic analysis reported seismic loads on the MHM uplift restraints and EDERER performed corresponding structural calculations to demonstrate structural adequacy of the seismic uplift restraint hardware. The ALSTHOM and EDERER calculations were performed for a parked MHM with the active seismic restraints engaged, resulting in uplift restraint loading only in the vertical direction. 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 active seismic restraints are not engaged. If a seismic event occurs during MHM repositioning operations, a moving contact at a seismic uplift restraint would introduce a friction load on the restraint in the direction of the movement. These potential horizontal friction loads on the uplift restraints were not included in the existing restraint hardware design calculations. One of the purposes of the current evaluation is to address the structural adequacy of the MHM seismic uplift restraints with the addition of the horizontal friction associated with MHM repositioning movements

76 FR 57767 - Proposed Generic Communication; Draft NRC Generic Letter 2011-XX: Seismic Risk Evaluations for...

Science.gov (United States)

2011-09-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0204] Proposed Generic Communication; Draft NRC Generic Letter 2011-XX: Seismic Risk Evaluations for Operating Reactors AGENCY: Nuclear Regulatory Commission... FR 54507), that requested public comment on Draft NRC Generic Letter 2011- XX: Seismic Risk...

Seismic shear wall ISP NUPEC's seismic ultimate dynamic response test. Comparison report

International Nuclear Information System (INIS)

1996-01-01

In the seismic design of a nuclear power plant, evaluation of the ultimate strength of the nuclear reactor building is an important subject for assessment of seismic reliability of the plant. In order to carry out the evaluation, the response characteristics of reinforced concrete seismic shear walls up to their ultimate state have to be understood. For this purpose, there is a need to develop reliable non-linear response analysis methods which enables the reliable ultimate strength evaluation of nuclear reactor buildings. Along with this need, many computer codes have been developed. These computer codes are compared. (K.A.)

The ESI scale, an ethical approach to the evaluation of seismic hazards

Science.gov (United States)

Porfido, Sabina; Nappi, Rosa; De Lucia, Maddalena; Gaudiosi, Germana; Alessio, Giuliana; Guerrieri, Luca

2015-04-01

The dissemination of correct information about seismic hazard is an ethical duty of scientific community worldwide. A proper assessment of a earthquake severity and impact should not ignore the evaluation of its intensity, taking into account both the effects on humans, man-made structures, as well as on the natural evironment. We illustrate the new macroseismic scale that measures the intensity taking into account the effects of earthquakes on the environment: the ESI 2007 (Environmental Seismic Intensity) scale (Michetti et al., 2007), ratified by the INQUA (International Union for Quaternary Research) during the XVII Congress in Cairns (Australia). The ESI scale integrates and completes the traditional macroseismic scales, of which it represents the evolution, allowing to assess the intensity parameter also where buildings are absent or damage-based diagnostic elements saturate. Each degree reflects the corresponding strength of an earthquake and the role of ground effects, evaluating the Intensity on the basis of the characteristics and size of primary (e.g. surface faulting and tectonic uplift/subsidence) and secondary effects (e.g. ground cracks, slope movements, liquefaction phenomena, hydrological changes, anomalous waves, tsunamis, trees shaking, dust clouds and jumping stones). This approach can be considered "ethical" because helps to define the real scenario of an earthquake, regardless of the country's socio-economic conditions and level of development. Here lies the value and the relevance of macroseismic scales even today, one hundred years after the death of Giuseppe Mercalli, who conceived the homonymous scale for the evaluation of earthquake intensity. For an appropriate mitigation strategy in seismic areas, it is fundamental to consider the role played by seismically induced effects on ground, such as active faults (size in length and displacement) and secondary effects (the total area affecting). With these perspectives two different cases

Seismic hazards: New trends in analysis using geologic data

International Nuclear Information System (INIS)

Schwartz, D.P.; Coppersmith, K.J.

1986-01-01

In the late 1960s and early 1970s, largely in response to expansion of nuclear power plant siting and issuance of a code of federal regullations by the Nuclear Regulatory Commission referred to as Appendix A-10CFR100, the need to characterize the earthquake potential of individual faults for seismic design took on greater importance. Appendix A established deterministic procedures for assessing the seismic hazard at nuclear power plant sites. Bonilla and Buchanan, using data from historical suface-faulting earthquakes, developed a set of statistical correlations relating earthquake magnitude to surface rupture length and to surface displacement. These relationships have been refined and updated along with the relationship between fault area and magnitude and seismic moment and moment magnitude have served as the basis for selecting maximum earthquakes in a wide variety of design situations. In the paper presented, the authors discuss new trends in seismic hazard analysis using geologic data, with special emphasis on fault-zone segmentation and recurrence models and the way in which they provide a basis for evaluating long-term earthquake potential

Safety review for seismic qualification on nuclear power plant equipment

International Nuclear Information System (INIS)

Fang Qingxian

1995-01-01

The standards and requirements for seismic qualification of nuclear power plant's component have been fully addressed, including the scope of seismic qualification, the approach and the method of common seismic qualification, the procedure of the seismic tests, and the criteria for the seismic qualification review. The problems discovered in the safety review and the solution for these problems and some other issues are also discussed

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

Evaluation of Seismic Behaviors of Partially Deteriorated Reinforced Concrete Circular Columns Retrofitted with CFRP

Directory of Open Access Journals (Sweden)

Dongxu Hou

2014-01-01

Full Text Available Deficiency of the concrete strength in some regions of reinforced concrete (RC columns in practice may weaken the seismic behaviors of columns. Its effects on RC columns should be well understood. This paper aims to investigate the influences of deteriorated segment on the seismic behaviors of partially deteriorated RC columns and attempts to recover the seismic behaviors of partially deteriorated columns with Carbon Fiber Reinforced Polymer (CFRP composites. A finite element analysis was carried out to simulate the seismic behaviors of CFRP-confined partially deteriorated RC columns. The numerical results were verified by the laboratory tests of six specimens. Based on the finite element results, the failure location of partially deteriorated columns in an earthquake was predicted, and the effectiveness of CFRP retrofitted on partially deteriorated columns was evaluated.

Recent achievements of the neo-deterministic seismic hazard assessment in the CEI region

International Nuclear Information System (INIS)

Panza, G.F.; Vaccari, F.; Kouteva, M.

2008-03-01

A review of the recent achievements of the innovative neo-deterministic approach for seismic hazard assessment through realistic earthquake scenarios has been performed. The procedure provides strong ground motion parameters for the purpose of earthquake engineering, based on the deterministic seismic wave propagation modelling at different scales - regional, national and metropolitan. The main advantage of this neo-deterministic procedure is the simultaneous treatment of the contribution of the earthquake source and seismic wave propagation media to the strong motion at the target site/region, as required by basic physical principles. The neo-deterministic seismic microzonation procedure has been successfully applied to numerous metropolitan areas all over the world in the framework of several international projects. In this study some examples focused on CEI region concerning both regional seismic hazard assessment and seismic microzonation of the selected metropolitan areas are shown. (author)

Seismic evaluation of vulnerability for SAMA educational buildings in Tehran

International Nuclear Information System (INIS)

Amini, Omid Nassiri; Amiri, Javad Vaseghi

2008-01-01

Earthquake is a destructive phenomenon that trembles different parts of the earth yearly and causes many destructions. Iran is one of the (high seismicity) quack- prone parts of the world that has received a lot of pecuniary damages and life losses each year, schools are of the most important places to be protected during such crisis.There was no special surveillance on designing and building of school's building in Tehran till the late 70's, and as Tehran is on faults, instability of such buildings may cause irrecoverable pecuniary damages and especially life losses, therefore preventing this phenomenon is in an urgent need.For this purpose, some of the schools built during 67-78 mostly with Steel braced frame structures have been selected, first, by evaluating the selected Samples, gathering information and Visual Survey, the prepared questionnaires were filled out. With the use of ARIA and SABA (Venezuela) Methods, new modified combined method for qualified evaluations was found and used.Then, for quantified evaluation, with the use of computer 3D models and nonlinear statically analysis methods, a number of selected buildings of qualified evaluation, were reevaluated and finally with nonlinear dynamic analysis method the real behavior of structures on the earthquakes is studied.The results of qualified and quantified evaluations were compared and a proper Pattern for seismic evaluation of Educational buildings was presented. Otherwise the results can be a guidance for the person in charge of retrofitting or if necessary rebuilding the schools

Seismic safety margin assessment program (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

2011-01-01

Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

Development of seismic hazard analysis in Japan

International Nuclear Information System (INIS)

Itoh, T.; Ishii, K.; Ishikawa, Y.; Okumura, T.

1987-01-01

In recent years, seismic risk assessment of the nuclear power plant have been conducted increasingly in various countries, particularly in the United States to evaluate probabilistically the safety of existing plants under earthquake loading. The first step of the seismic risk assessment is the seismic hazard analysis, in which the relationship between the maximum earthquake ground motions at the plant site and their annual probability of exceedance, i.e. the seismic hazard curve, is estimated. In this paper, seismic hazard curves are evaluated and examined based on historical earthquake records model, in which seismic sources are modeled with area-sources, for several different sites in Japan. A new evaluation method is also proposed to compute the response spectra of the earthquake ground motions in connection with estimating the probabilistic structural response. Finally the numerical result of probabilistic risk assessment for a base-isolated three story RC structure, in which the frequency of seismic induced structural failure is evaluated combining the seismic hazard analysis, is described briefly

The development of seismic guidelines for the Stanford Linear Accelerator Center

International Nuclear Information System (INIS)

Huggins, R.

1996-08-01

This paper describes the development of Seismic Guidelines for the Stanford Linear Accelerator Center (SLAC). Although structures have always been built conservatively, SLAC management decided to review and update their seismic guidelines. SLAC is about mid-way between the epicenters of the 8.3 Richter magnitude 1906 San Francisco and the 7.2 Loma Prieta Earthquakes. The west end of the two mile long electron/positron particle accelerator lies a half mile from the large San Andreas Fault. Suggestions for seismic planning processes were solicited from local computer manufacturing firms, universities, and federal laboratories. A Committee of the various stakeholders in SLAC's seismic planning retained an internationally known Seismic Planning Consultant and reviewed relevant standards and drafted Guidelines. A panel of seismic experts was convened to help define the hazard, site response spectra, probabilistic analysis of shaking, and near field effects. The Facility's structures were assigned to seismic classes of importance, and an initial assessment of a sample of a dozen buildings conducted. This assessment resulted in emergency repairs to one structure, and provided a open-quotes reality basisclose quotes for establishing the final Guidelines and Administrative Procedures, and a program to evaluate remaining buildings, shielding walls, tunnels, and other special structures

The development of seismic guidelines for the Stanford Linear Accelerator Center

Energy Technology Data Exchange (ETDEWEB)

Huggins, R.

1996-08-01

This paper describes the development of Seismic Guidelines for the Stanford Linear Accelerator Center (SLAC). Although structures have always been built conservatively, SLAC management decided to review and update their seismic guidelines. SLAC is about mid-way between the epicenters of the 8.3 Richter magnitude 1906 San Francisco and the 7.2 Loma Prieta Earthquakes. The west end of the two mile long electron/positron particle accelerator lies a half mile from the large San Andreas Fault. Suggestions for seismic planning processes were solicited from local computer manufacturing firms, universities, and federal laboratories. A Committee of the various stakeholders in SLAC`s seismic planning retained an internationally known Seismic Planning Consultant and reviewed relevant standards and drafted Guidelines. A panel of seismic experts was convened to help define the hazard, site response spectra, probabilistic analysis of shaking, and near field effects. The Facility`s structures were assigned to seismic classes of importance, and an initial assessment of a sample of a dozen buildings conducted. This assessment resulted in emergency repairs to one structure, and provided a {open_quotes}reality basis{close_quotes} for establishing the final Guidelines and Administrative Procedures, and a program to evaluate remaining buildings, shielding walls, tunnels, and other special structures.

Improved Seismic Acquisition System and Data Processing for the Italian National Seismic Network

Science.gov (United States)

Badiali, L.; Marcocci, C.; Mele, F.; Piscini, A.

2001-12-01

A new system for acquiring and processing digital signals has been developed in the last few years at the Istituto Nazionale di Geofisica e Vulcanologia (INGV). The system makes extensive use of the internet communication protocol standards such as TCP and UDP which are used as the transport highway inside the Italian network, and possibly in a near future outside, to share or redirect data among processes. The Italian National Seismic Network has been working for about 18 years equipped with vertical short period seismometers and transmitting through analog lines, to the computer center in Rome. We are now concentrating our efforts on speeding the migration towards a fully digital network based on about 150 stations equipped with either broad band or 5 seconds sensors connected to the data center partly through wired digital communication and partly through satellite digital communication. The overall process is layered through intranet and/or internet. Every layer gathers data in a simple format and provides data in a processed format, ready to be distributed towards the next layer. The lowest level acquires seismic data (raw waveforms) coming from the remote stations. It handshakes, checks and sends data in LAN or WAN according to a distribution list where other machines with their programs are waiting for. At the next level there are the picking procedures, or "pickers", on a per instrument basis, looking for phases. A picker spreads phases, again through the LAN or WAN and according to a distribution list, to one or more waiting locating machines tuned to generate a seismic event. The event locating procedure itself, the higher level in this stack, can exchange information with other similar procedures. Such a layered and distributed structure with nearby targets allows other seismic networks to join the processing and data collection of the same ongoing event, creating a virtual network larger than the original one. At present we plan to cooperate with other

Effective updating process of seismic fragilities using Bayesian method and information entropy

International Nuclear Information System (INIS)

Kato, Masaaki; Takata, Takashi; Yamaguchi, Akira

2008-01-01

Seismic probabilistic safety assessment (SPSA) is an effective method for evaluating overall performance of seismic safety of a plant. Seismic fragilities are estimated to quantify the seismically induced accident sequences. It is a great concern that the SPSA results involve uncertainties, a part of which comes from the uncertainty in the seismic fragility of equipment and systems. A straightforward approach to reduce the uncertainty is to perform a seismic qualification test and to reflect the results on the seismic fragility estimate. In this paper, we propose a figure-of-merit to find the most cost-effective condition of the seismic qualification tests about the acceleration level and number of components tested. Then a mathematical method to reflect the test results on the fragility update is developed. A Bayesian method is used for the fragility update procedure. Since a lognormal distribution that is used for the fragility model does not have a Bayes conjugate function, a parameterization method is proposed so that the posterior distribution expresses the characteristics of the fragility. The information entropy is used as the figure-of-merit to express importance of obtained evidence. It is found that the information entropy is strongly associated with the uncertainty of the fragility. (author)

Simulating the seismic behaviour of soil slopes and embankments

DEFF Research Database (Denmark)

Zania, Varvara; Tsompanakis, Yiannis; Psarropoulos, Prodromos

2010-01-01

In the current study the clarification of the main assumptions, related to the two most commonly used methods of seismic slope stability analysis (pseudostatic and permanent deformation) is attempted. The seismic permanent displacements and the corresponding seismic coefficients were determined via...... parametric dynamic numerical analyses taking into account not only the main parameters dominating the seismic slope stability, but also the inherent assumptions of the applied approaches that affect the obtained results. The investigation conclude to a realistic procedure for seismic slope stability...

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

Calibration of Seismic Attributes for Reservoir Characterization

Energy Technology Data Exchange (ETDEWEB)

Wayne D. Pennington

2002-09-29

The project, "Calibration of Seismic Attributes for Reservoir Characterization," is now complete. Our original proposed scope of work included detailed analysis of seismic and other data from two to three hydrocarbon fields; we have analyzed data from four fields at this level of detail, two additional fields with less detail, and one other 2D seismic line used for experimentation. We also included time-lapse seismic data with ocean-bottom cable recordings in addition to the originally proposed static field data. A large number of publications and presentations have resulted from this work, inlcuding several that are in final stages of preparation or printing; one of these is a chapter on "Reservoir Geophysics" for the new Petroleum Engineering Handbook from the Society of Petroleum Engineers. Major results from this project include a new approach to evaluating seismic attributes in time-lapse monitoring studies, evaluation of pitfalls in the use of point-based measurements and facies classifications, novel applications of inversion results, improved methods of tying seismic data to the wellbore, and a comparison of methods used to detect pressure compartments. Some of the data sets used are in the public domain, allowing other investigators to test our techniques or to improve upon them using the same data. From the public-domain Stratton data set we have demonstrated that an apparent correlation between attributes derived along 'phantom' horizons are artifacts of isopach changes; only if the interpreter understands that the interpretation is based on this correlation with bed thickening or thinning, can reliable interpretations of channel horizons and facies be made. From the public-domain Boonsville data set we developed techniques to use conventional seismic attributes, including seismic facies generated under various neural network procedures, to subdivide regional facies determined from logs into productive and non-productive subfacies, and we

CALIBRATION OF SEISMIC ATTRIBUTES FOR RESERVOIR CHARACTERIZATION

Energy Technology Data Exchange (ETDEWEB)

Wayne D. Pennington; Horacio Acevedo; Aaron Green; Joshua Haataja; Shawn Len; Anastasia Minaeva; Deyi Xie

2002-10-01

The project, ''Calibration of Seismic Attributes for Reservoir Calibration,'' is now complete. Our original proposed scope of work included detailed analysis of seismic and other data from two to three hydrocarbon fields; we have analyzed data from four fields at this level of detail, two additional fields with less detail, and one other 2D seismic line used for experimentation. We also included time-lapse seismic data with ocean-bottom cable recordings in addition to the originally proposed static field data. A large number of publications and presentations have resulted from this work, including several that are in final stages of preparation or printing; one of these is a chapter on ''Reservoir Geophysics'' for the new Petroleum Engineering Handbook from the Society of Petroleum Engineers. Major results from this project include a new approach to evaluating seismic attributes in time-lapse monitoring studies, evaluation of pitfalls in the use of point-based measurements and facies classifications, novel applications of inversion results, improved methods of tying seismic data to the wellbore, and a comparison of methods used to detect pressure compartments. Some of the data sets used are in the public domain, allowing other investigators to test our techniques or to improve upon them using the same data. From the public-domain Stratton data set we have demonstrated that an apparent correlation between attributes derived along ''phantom'' horizons are artifacts of isopach changes; only if the interpreter understands that the interpretation is based on this correlation with bed thickening or thinning, can reliable interpretations of channel horizons and facies be made. From the public-domain Boonsville data set we developed techniques to use conventional seismic attributes, including seismic facies generated under various neural network procedures, to subdivide regional facies determined from logs into

Development of seismic damage assessment system for nuclear power plant structures in Korea

International Nuclear Information System (INIS)

Hyun, Chang-Hun; Lee, Sung-Kyu; Choi, Kang-Ryoung; Koh, Hyun-Moo; Cho, HoHyun

2003-01-01

A seismic damage assessment system that analyses in real-time the actual seismic resistance capacity and the damage level of power plant structures has been developed. The system consists of three parts: a 3-D inelastic seismic analysis, a damage assessment using a damage index based on the previous 3-D analysis, and a 3-D graphic representation. PSC containment structures are modelled by finite shell elements using layered method and analysis is performed by means of time history inelastic seismic analysis method, which takes into account material nonlinearities. HHT-α, one kind of direct integration method, is adopted for the seismic analysis. Two damage indices at finite element and structural levels are applied for the seismic damage assessment. 3-D graphical representation of dynamic responses and damage index expedites procedure for evaluating the damage level. The developed system is now being installed at the Earthquake Monitoring Center of KINS (Korea Institute of Nuclear Safety) to support site inspections after an earthquake occurrence, and decisions about effective emergency measures, repair and operations of the plant. (author)

Seismic fragility capacity of equipment

International Nuclear Information System (INIS)

Iijima, Toru; Abe, Hiroshi; Suzuki, Kenichi

2006-01-01

Seismic probabilistic safety assessment (PSA) is an available method to evaluate residual risks of nuclear plants that are designed on definitive seismic conditions. From our preliminary seismic PSA analysis, horizontal shaft pumps are important components that have significant influences on the core damage frequency (CDF). An actual horizontal shaft pump and some kinds of elements were tested to evaluate realistic fragility capacities. Our test results showed that the realistic fragility capacity of horizontal shaft pump would be at least four times as high as a current value, 1.6 x 9.8 m/s 2 , used for our seismic PSA. We are going to incorporate the fragility capacity data that were obtained from those tests into our seismic PSA analysis, and we expect that the reliability of seismic PSA should increase. (author)

CARES-ESTSC, Seismic Structure Safety Analysis for Nuclear Power Plants

International Nuclear Information System (INIS)

Costantino, C.J.; Miller, C.A.; Heymsfield, E.; Yang, A.

1999-01-01

1 - Description of program or function: CARES, Computer Analysis for Rapid Evaluation of Structures, was developed for NRC staff use to determine the validity and accuracy of the analysis methods used by various utilities for structural safety evaluations of nuclear power plants. CARES is organized in a modular format with the basic modules of the system performing static, seismic, and nonlinear analysis. In this release, only the seismic module is implemented. This module defines the design seismic criteria at a given site, evaluates the free-field motion, and computes the structural response and floor response spectra including soil-structure interaction. The eight options in CARES currently are: a general manager for the seismic module, deconvolution analysis, structural data preparation for soil-structure interaction (SSI) analysis, input motion preparation for SSI analysis, SSI analysis, earthquake simulations/data, PSD (Power Spectral Density) related acceleration time history/spectra analysis, and plot generation. 2 - Method of solution: The seismic module works in the frequency domain. Earthquake motion simulation is based on the fundamental property that any periodic function can be expanded in a series of sinusoidal waves. The computer uses a random number generator to produce strings of phase angles with uniform distribution in the 0-2 pi range. Then, a linear correction procedure due to Scanlon and Sacks is employed to derive an adjusted array of amplitudes. The acceleration ensemble is subsequently modified by a deterministic intensity function composed of three segments: an initial buildup, a stationary duration, and exponential steady decay. A parabolic correction procedure outlined by Jennings and Housner is applied to the acceleration ensemble to bring the end velocity of the ground motion to zero. The soil-structure system is represented by a three-dimensional lumped parameter type model. The structural model is built up from three

Seismic verification of nuclear plant equipment anchorage

International Nuclear Information System (INIS)

Lepiece, M.; Van Vyve, J.

1991-01-01

More than 60% of the electrical power of Belgium is generated by seven PWR nuclear power plants. For three of them, the electro-mechanical equipment had to be reassessed after ten years of operation, because the seismic requirements were upgraded from 0.1 g to 0.17 g free field ground acceleration. The seismic requalification of the active equipment was a critical problem as the classical methods were too conservative. The approach based on the use of the past experience on the seismic behavior of nonnuclear equipment, chosen and developed by the SQUG, had to be transposed to the Belgian N.P.P. The decision of the accept-ability of equipment relies heavily on the aseismatic capacity of anchorage. The Electrical Power Research Institute (EPRI) developed the procedure and guideline for the demonstration of the aseismatic adequacy of equipment anchorage in a cost-effective and consistent manner, to support the decision by Seismic Review Team. The field inspection procedure to identify the type of fasteners and detect their possible defects and the verification procedure developed to calculate the aseismatic capacity of equipment anchorage on the strength of fasteners, the aseismatic capacity of anchorage and the comparison of the capacity with the demand are reported. (K.I.)

Containment performance evaluation for the GESSAR-II plant for seismic initiating events

International Nuclear Information System (INIS)

Shiu, K.K.; Chu, T.; Ludewig, H.; Pratt, W.T.

1986-01-01

As a part of the overall effort undertaken by Brookhaven National Laboratory (BNL) to review the GESSAR-II probabilistic risk assessment, an independent containment performance evaluation was performed using the containment event tree approach. This evaluation focused principally on those accident sequences which are initiated by seismic events. This paper reports the findings of this study. 1 ref

Role of seismic PRA in seismic safety decisions of nuclear power plants

International Nuclear Information System (INIS)

Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

1985-01-01

This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

Evaluation of seismic stability of near field

International Nuclear Information System (INIS)

Taniguchi, Wataru; Takaji, Kazuhiko; Sugino, Hiroyuki; Mori, Koji

1999-11-01

For the buffer material of geological disposal of high-level radioactive waste (HLW) in Japan, it is considered to use a compacted bentonite or a compacted sand-mixture bentonite that is one kind of clay. The buffer material is expected to maintain long-term mechanical stability, to hold the waste in designated place, and to avoid the effects on the radionuclides migration. It is considered that the cyclic load due to seismic activities affects long-term mechanical stability in Japan, where many earthquakes have been occurring. In this report, aseismic mechanical stability of engineered barrier of HLW is studied by dynamic analysis based on equation of vibration, mainly in the view point of mechanical stability of the buffer. The analytical computer code that has been developed by JNC in cooperative project with National Research Institute for Earth Science and Disaster Prevention Science and Technology Agency is used in this study. Seismic wave at the disposal depth in the assumed geological environment is established by multiple reflection theory analysis, and then seismic wave at the disposal depth is used for the aseismic mechanical stability analysis. For the aseismic mechanical stability, total stress analyses (single-phase system) with the target field of near field are conducted to evaluate the shear failure of the buffer, the displacement of overpack, and vibrational behavior of the engineered barrier, and then effective stress analyses (two-phase system) with the target field of the engineered barrier are conducted to evaluate excursion in the pore water pressure within the buffer (i. e. liquefaction), concerning the non-linear dynamic properties of the buffer material. From the results, the following conclusions are obtained. (1) From the results of the total stress analyses, it is confirmed that the buffer must not reach a shear failure condition from the stresses caused by an earthquake and the overpack must not move significantly due to the inertial

Seismic evaluation of a cooling water reservoir facility including fluid-structure and soil-structure interaction effects

International Nuclear Information System (INIS)

Kabir, A.F.; Maryak, M.E.

1991-01-01

Seismic analyses and structural evaluations were performed for a cooling water reservoir of a nuclear reactor facility. The horizontal input seismic motion was the NRC Reg. guide 1.60 spectrum shape anchored at 0.20g zero period acceleration. Vertical input was taken as two-thirds of the horizontal input. Soil structure interaction and hydrodynamic effects were addressed in the seismic analyses. Uncertainties in the soil properties were accounted for by considering three soil profiles. Two 2-dimensional SSI models and a 3-dimensional static model. Representing different areas of the reservoir structures were developed and analyzed to obtain seismic forces and moments, and accelerations at various locations. The results included in this paper indicated that both hydrodynamic and soil-structure interaction effects are significant contributors to the seismic responses of the water-retaining walls of the reservoir

Research items regarding seismic residual risk evaluation

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

After learning the Fukushima Dai-ichi NPP severe accidents in 2011, the government investigation committee proposed the effective use of probabilistic safety assessment (PSA), and now it is required to establish new safety rules reflecting the results of probabilistic risk assessment (PRA) and proposed severe accident measures. Since the Seismic Design Guide has been revised on September 19, 2006, JNES has been discussing seismic PRA (Levels 1-3) methods to review licensees' residual risk assessment while preparing seismic PRA models. Meanwhile, new safety standards for light water reactors are to be issued and enforced on July 2013, which require the residual risk of tsunami, in addition to earthquakes, should be lowered as much as possible. The Fukushima accidents raised the problems related to risk assessment, e.g. approaches based on multi-hazard (earthquake and tsunami), multi-unit, multi-site, and equipment's common cause failure. This fiscal year, while performing seismic and/or tsunami PRA to work on these problems, JNES picked up the equipment whose failure greatly contribute to core damage, surveyed accident management measures on those equipment as well as effectiveness to reduce core damage probability. (author)

The New Italian Seismic Hazard Model

Science.gov (United States)

Marzocchi, W.; Meletti, C.; Albarello, D.; D'Amico, V.; Luzi, L.; Martinelli, F.; Pace, B.; Pignone, M.; Rovida, A.; Visini, F.

2017-12-01

In 2015 the Seismic Hazard Center (Centro Pericolosità Sismica - CPS) of the National Institute of Geophysics and Volcanology was commissioned of coordinating the national scientific community with the aim to elaborate a new reference seismic hazard model, mainly finalized to the update of seismic code. The CPS designed a roadmap for releasing within three years a significantly renewed PSHA model, with regard both to the updated input elements and to the strategies to be followed. The main requirements of the model were discussed in meetings with the experts on earthquake engineering that then will participate to the revision of the building code. The activities were organized in 6 tasks: program coordination, input data, seismicity models, ground motion predictive equations (GMPEs), computation and rendering, testing. The input data task has been selecting the most updated information about seismicity (historical and instrumental), seismogenic faults, and deformation (both from seismicity and geodetic data). The seismicity models have been elaborating in terms of classic source areas, fault sources and gridded seismicity based on different approaches. The GMPEs task has selected the most recent models accounting for their tectonic suitability and forecasting performance. The testing phase has been planned to design statistical procedures to test with the available data the whole seismic hazard models, and single components such as the seismicity models and the GMPEs. In this talk we show some preliminary results, summarize the overall strategy for building the new Italian PSHA model, and discuss in detail important novelties that we put forward. Specifically, we adopt a new formal probabilistic framework to interpret the outcomes of the model and to test it meaningfully; this requires a proper definition and characterization of both aleatory variability and epistemic uncertainty that we accomplish through an ensemble modeling strategy. We use a weighting scheme

Simulation-based seismic loss estimation of seaport transportation system

International Nuclear Information System (INIS)

Ung Jin Na; Shinozuka, Masanobu

2009-01-01

Seaport transportation system is one of the major lifeline systems in modern society and its reliable operation is crucial for the well-being of the public. However, past experiences showed that earthquake damage to port components can severely disrupt terminal operation, and thus negatively impact on the regional economy. The main purpose of this study is to provide a methodology for estimating the effects of the earthquake on the performance of the operation system of a container terminal in seaports. To evaluate the economic loss of damaged system, an analytical framework is developed by integrating simulation models for terminal operation and fragility curves of port components in the context of seismic risk analysis. For this purpose, computerized simulation model is developed and verified with actual terminal operation records. Based on the analytical procedure to assess the seismic performance of the terminal, system fragility curves are also developed. This simulation-based loss estimation methodology can be used not only for estimating the seismically induced revenue loss but also serve as a decision-making tool to select specific seismic retrofit technique on the basis of benefit-cost analysis

Test and evaluation about damping characteristics of hanger supports for nuclear power plant piping systems (Seismic Damping Ratio Evaluation Program)

International Nuclear Information System (INIS)

Shibata, H.; Ito, A.; Tanaka, K.; Niino, T.; Gotoh, N.

1981-01-01

Generally, damping phenomena of structures and equipments is caused by very complex energy dissipation. Especially, as piping systems are composed of many components, it is very difficult to evaluate damping characteristics of its system theoretically. On the other hand, the damping value for aseismic design of nuclear power plants is very important design factor to decide seismic response loads of structures, equipments and piping systems. The very extensive studies titled SDREP (Seismic Damping Ratio Evaluation Program) were performed to establish proper damping values for seismic design of piping as a joint work among a university, electric companies and plant makers. In SDREP, various systematic vibration tests were conducted to investigate factors which may contribute to damping characteristics of piping systems and to supplement the data of the pre-operating tests. This study is related to the component damping characteristics tests of that program. The object of this study is to clarify damping characteristics and mechanism of hanger supports used in piping systems, and to establish the evaluation technique of dispersing energy at hanger support points and its effect to the total damping ability of piping system. (orig./WL)

CARES (Computer Analysis for Rapid Evaluation of Structures) Version 1.0, seismic module

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulas, A.J.; Miller, C.A.; Costantino, C.J.

1990-07-01

During FY's 1988 and 1989, Brookhaven National Laboratory (BNL) developed the CARES system (Computer Analysis for Rapid Evaluation of Structures) for the US Nuclear Regulatory Commission (NRC). CARES is a PC software system which has been designed to perform structural response computations similar to those encountered in licensing reviews of nuclear power plant structures. The documentation of the Seismic Module of CARES consists of three volumes. This report is Volume 2 of the three volume documentation of the Seismic Module of CARES and represents the User's Manual. 14 refs

GIS-based seismic shaking slope vulnerability map of Sicily (Central Mediterranean)

Science.gov (United States)

Nigro, Fabrizio; Arisco, Giuseppe; Perricone, Marcella; Renda, Pietro; Favara, Rocco

2010-05-01

permanent displacement potentially induced by an seismic scenario. Such methodologies found on the consideration that the conditions of seismic stability and the post-seismic functionality of engineering structures are tightly related to the entity of the permanent deformations that an earthquake can induce. Regarding the existing simplified procedures among slope stability models, Newmark's model is often used to derive indications about slope instabilities due to earthquakes. In this way, we have evaluated the seismically-induced landslides hazard in Sicily (Central Mediterranean) using the Newmark-like model. In order to determine the map distribution of the seismic ground-acceleration from an earthquake scenario, the attenuation-law of Sabetta & Pugliese has been used, analyzing some seismic recordings occurred in Italy. Also, by evaluating permanent displacements, the correlation of Ambraseys & Menu has been assumed. The seismic shaking slope vulnerability map of Sicily has been carried out using GIS application, also considering max seismic ground-acceleration peak distribution (in terms of exceedance probability for fixed time), slope acclivity, cohesion/angle of internal friction of outcropping rocks, allowing the zoning of the unstable slopes under seismic forces.

Is Fuel Assembly Fine at BDBA Seismic Load?

Energy Technology Data Exchange (ETDEWEB)

Kang, Heung Seok; Lee, Kang Hee; Yoon, Kyung Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

After Fukushima accident, IAEA and OECD/NEA speak aloud recommendation on Design Extension Condition (DEC) for some of current BDBA accidents, and thus, some of the current BDBA to be obviously included in design conditions. In this study, 1) we will review on 2011 Fukushima accident from the earthquake point of view, before great tsunami, 2) on the analysis procedure for seismic accidents, of which the main frame was established several decades ago, 3) on possible issue on current design method, and 4) on practical way to solve the design issues and to reflect a beyond design basis seismic accident in DEC. In this study, we have reviewed seismic analysis procedure and tests for FA mechanical integrity. We may give some recommendation to incorporate BDB seismic accident into DEC as follows: 1) FA characteristic test considering realistic boundary conditions 2) Implementation of FSI into analysis models 3) Verification test to confirm design and safety margin.

Evaluation of Fourier and Response Spectra at Ichihasama and Koromogawa Seismic Intensity Observation Sites During the Iwate-Miyagi Nairiku Earthquake in 2008

Science.gov (United States)

Nishikawa, Hayato; Miyajima, Masakatsu

In this study, we evaluate an acceleration Fourier and response spectra at Ichihasama and Koromogawa seismic intensity observation sites which observed JMA seismic intensity of 6 upper but seismic waveform records don't exist during the Iwate-Miyagi Nairiku earthquake in 2008. Firstly, formula to evaluate acceleration Fourier and response spectra are developed using peak ground acceleration, JMA seismic intensity and predominant period of earthquake spectra based on records obtained from crustal earthquakes with Magnitude of 6 to 7. Acceleration Fourier and response spectra are evaluated for another local government site which are not chosen for development of the formula. The evaluated values mostly agree with the observed ones. Finally, acceleration Fourier and response spectra are evaluated for Ichihasama and Koromogawa observation sites. It is clarified that short period below 1 second was predominated in the evaluated spectra.

Seismic proving test of process computer systems with a seismic floor isolation system

International Nuclear Information System (INIS)

Fujimoto, S.; Niwa, H.; Kondo, H.

1995-01-01

The authors have carried out seismic proving tests for process computer systems as a Nuclear Power Engineering Corporation (NUPEC) project sponsored by the Ministry of International Trade and Industry (MITI). This paper presents the seismic test results for evaluating functional capabilities of process computer systems with a seismic floor isolation system. The seismic floor isolation system to isolate the horizontal motion was composed of a floor frame (13 m x 13 m), ball bearing units, and spring-damper units. A series of seismic excitation tests was carried out using a large-scale shaking table of NUPEC. From the test results, the functional capabilities during large earthquakes of computer systems with a seismic floor isolation system were verified

Probabilistic seismic vulnerability and risk assessment of stone masonry structures

Science.gov (United States)

Abo El Ezz, Ahmad

Earthquakes represent major natural hazards that regularly impact the built environment in seismic prone areas worldwide and cause considerable social and economic losses. The high losses incurred following the past destructive earthquakes promoted the need for assessment of the seismic vulnerability and risk of the existing buildings. Many historic buildings in the old urban centers in Eastern Canada such as Old Quebec City are built of stone masonry and represent un-measurable architectural and cultural heritage. These buildings were built to resist gravity loads only and generally offer poor resistance to lateral seismic loads. Seismic vulnerability assessment of stone masonry buildings is therefore the first necessary step in developing seismic retrofitting and pre-disaster mitigation plans. The objective of this study is to develop a set of probability-based analytical tools for efficient seismic vulnerability and uncertainty analysis of stone masonry buildings. A simplified probabilistic analytical methodology for vulnerability modelling of stone masonry building with systematic treatment of uncertainties throughout the modelling process is developed in the first part of this study. Building capacity curves are developed using a simplified mechanical model. A displacement based procedure is used to develop damage state fragility functions in terms of spectral displacement response based on drift thresholds of stone masonry walls. A simplified probabilistic seismic demand analysis is proposed to capture the combined uncertainty in capacity and demand on fragility functions. In the second part, a robust analytical procedure for the development of seismic hazard compatible fragility and vulnerability functions is proposed. The results are given by sets of seismic hazard compatible vulnerability functions in terms of structure-independent intensity measure (e.g. spectral acceleration) that can be used for seismic risk analysis. The procedure is very efficient for

Seismic margins and calibration of piping systems

International Nuclear Information System (INIS)

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

1985-01-01

The Seismic Safety Margins Research Program (SSMRP) is a US Nuclear Regulatory Commission-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its objective is to develop a complete, fully coupled analysis procedure for estimating the risk of earthquake-induced radioactive release from a commercial nuclear power plant and to determine major contributors to the state-of-the-art seismic and systems analysis process and explicitly includes the uncertainties in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I of SSMRP, the overall seismic risk assessment methodology was developed and assembled. The application of this methodology to the seismic PRA (Probabilistic Risk Assessment) at the Zion Nuclear Power Plant has been documented. This report documents the method deriving response factors. The response factors, which relate design calculated responses to best estimate values, were used in the seismic response determination of piping systems for a simplified seismic probablistic risk assessment. 13 references, 31 figures, 25 tables

Seismic evaluation of K basin bridge cranes (HOI-320 ampersand HOI-418) and supporting structure

International Nuclear Information System (INIS)

Winkel, B.V.; Kanjilad, S.K.

1996-03-01

The Safety Class 1 100-K fuel storage basins are vulnerable to impact damage if a bridge crane were to fall during a seismic event. The pupose of this report is to address the adequacy of the K Basin bridge cranes to resist a seismic-induced fall. The approach used to demonstrate adequacy against falling, was to evaluate the crane structural components relative to requirements specified in ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes. Additionally, wheel lift-off and the adequacy of the crane supporting structure, are addressed. Seismic adequacy of the mechanical hoist equipment is not addressed in this report

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

Performance Evaluation of Submerged Floating Tunnel Subjected to Hydrodynamic and Seismic Excitations

Directory of Open Access Journals (Sweden)

Naik Muhammad

2017-10-01

Full Text Available Submerged floating tunnels (SFTs are innovative structural solutions to waterway crossings, such as sea-straits, fjords and lakes. As the width and depth of straits increase, the conventional structures such as cable-supported bridges, underground tunnels or immersed tunnels become uneconomical alternatives. For the realization of SFT, the structural response under extreme environmental conditions needs to be evaluated properly. This study evaluates the displacements and internal forces of SFT under hydrodynamic and three-dimensional seismic excitations to check the global performance of an SFT in order to conclude on the optimum design. The formulations incorporate modeling of ocean waves, currents and mooring cables. The SFT responses were evaluated using three different mooring cable arrangements to determine the stability of the mooring configuration, and the most promising configuration was then used for further investigations. A comparison of static, hydrodynamic and seismic response envelope curves of the SFT is provided to determine the dominant structural response. The study produces useful conclusions regarding the structural behavior of the SFT using a three-dimensional numerical model.

Seismic analysis of a nonlinear airlock system

International Nuclear Information System (INIS)

Huang, S.N.

1983-01-01

The containment equipment airlock door of the Fast Flux Test Facility utilizes screw-type actuators as a push-pull mechanism for closing and opening operations. Special design features were used to protect these actuators from pressure differential loading. These made the door behave as a nonlinear system during a seismic event. Seismic analyses, utilizing the time history method, were conducted to determine the seismic loads on these scew-type actuators. Several sizes of actuators were examined. Procedures for determining the final optimum design are discussed in detail

The influence of backfill on seismicity

CSIR Research Space (South Africa)

Hemp, DA

1990-09-01

Full Text Available , that the seismicity has been reduced in areas where backfill had been placed. A factor complicating the evaluation of backfill on seismicity is the effect of geological structures on seismicity....

New Evaluation of Seismic Hazard in Cental America and la Hispaniola

Science.gov (United States)

Benito, B.; Camacho, E. I.; Rojas, W.; Climent, A.; Alvarado-Induni, G.; Marroquin, G.; Molina, E.; Talavera, E.; Belizaire, D.; Pierristal, G.; Torres, Y.; Huerfano, V.; Polanco, E.; García, R.; Zevallos, F.

2013-05-01

The results from seismic hazard studies carried out in two seismic scenarios, Central America Region (CA) and La Hispaniola Island, are presented here. Both cases follow the Probabilistic Seismic Hazard Assessment (PSHA) methodology and they are developed in terms of PGA, and SA (T), for T of 0.1, 0.2, 0.5, 1 and 2s. In both anaysis, hybrid zonation models are considered, integrated by seismogenic zones and faults where data of slip rate and recurrence time are available. First, we present a new evaluation of seismic hazard in CA, starting with the results of a previous study by Benito et al (2011). Some improvements are now included, such as: updated catalogue till 2011, corrections in the zonning model in particular for subduction regime taken into account the variation of the dip in Costa Rica and Panama, and modelization of some faults as independent units for the hazard estimation. The results allow us to carry out a sensitivity analysis comparing the ones obtained with and without faults. In a second part we present the results of the PSHA in La Hispaniola, carried out as part of the cooperative project SISMO-HAITI supported by UPM and developed in cooperation with ONEV. It started a few months after the 2010 event, as an answer to a required help from the Haitian government to UPM. The study was aimed at obtaining results suitable for seismic design purposes and started with the elaboration of a seismic catalogue for the Hispaniola, requiring an exhaustive revision of data reported by around 30 seismic agencies, apart from these from Puerto Rico and Dominican Republic Seismic Networks. Seismotectonic models for the region were reviewed and a new regional zonation was proposed, taking into account different geophysical data. Attenuation models for subduction and crustal zones were also reviewed and the more suitable were calibrated with data recorded inside the Caribbean plate. As a result of the PSHA, different maps were generated for the quoted parameters

Seismic hazard analysis. A methodology for the Eastern United States

Energy Technology Data Exchange (ETDEWEB)

Bernreuter, D L

1980-08-01

This report presents a probabilistic approach for estimating the seismic hazard in the Central and Eastern United States. The probabilistic model (Uniform Hazard Methodology) systematically incorporates the subjective opinion of several experts in the evaluation of seismic hazard. Subjective input, assumptions and associated hazard are kept separate for each expert so as to allow review and preserve diversity of opinion. The report is organized into five sections: Introduction, Methodology Comparison, Subjective Input, Uniform Hazard Methodology (UHM), and Uniform Hazard Spectrum. Section 2 Methodology Comparison, briefly describes the present approach and compares it with other available procedures. The remainder of the report focuses on the UHM. Specifically, Section 3 describes the elicitation of subjective input; Section 4 gives details of various mathematical models (earthquake source geometry, magnitude distribution, attenuation relationship) and how these models re combined to calculate seismic hazard. The lost section, Uniform Hazard Spectrum, highlights the main features of typical results. Specific results and sensitivity analyses are not presented in this report. (author)

An Ensemble Approach for Improved Short-to-Intermediate-Term Seismic Potential Evaluation

Science.gov (United States)

Yu, Huaizhong; Zhu, Qingyong; Zhou, Faren; Tian, Lei; Zhang, Yongxian

2017-06-01

Pattern informatics (PI), load/unload response ratio (LURR), state vector (SV), and accelerating moment release (AMR) are four previously unrelated subjects, which are sensitive, in varying ways, to the earthquake's source. Previous studies have indicated that the spatial extent of the stress perturbation caused by an earthquake scales with the moment of the event, allowing us to combine these methods for seismic hazard evaluation. The long-range earthquake forecasting method PI is applied to search for the seismic hotspots and identify the areas where large earthquake could be expected. And the LURR and SV methods are adopted to assess short-to-intermediate-term seismic potential in each of the critical regions derived from the PI hotspots, while the AMR method is used to provide us with asymptotic estimates of time and magnitude of the potential earthquakes. This new approach, by combining the LURR, SV and AMR methods with the choice of identified area of PI hotspots, is devised to augment current techniques for seismic hazard estimation. Using the approach, we tested the strong earthquakes occurred in Yunnan-Sichuan region, China between January 1, 2013 and December 31, 2014. We found that most of the large earthquakes, especially the earthquakes with magnitude greater than 6.0 occurred in the seismic hazard regions predicted. Similar results have been obtained in the prediction of annual earthquake tendency in Chinese mainland in 2014 and 2015. The studies evidenced that the ensemble approach could be a useful tool to detect short-to-intermediate-term precursory information of future large earthquakes.

Estimation of reservoir fluid volumes through 4-D seismic analysis on Gullfaks

Energy Technology Data Exchange (ETDEWEB)

Veire, H.S.; Reymond, S.B.; Signer, C.; Tenneboe, P.O.; Soenneland, L.; Schlumberger, Geco-Prakla

1998-12-31

4-D seismic has the potential to monitor hydrocarbon movement in reservoirs during production, and could thereby supplement the predictions of reservoir parameters offered by the reservoir simulator. However 4-D seismic is often more band limited than the vertical resolution required in the reservoir model. As a consequence the seismic data holds a composite response from reservoir parameter changes during production so that the inversion becomes non-unique. A procedure where data from the reservoir model are integrated with seismic data will be presented. The potential of such a procedure is demonstrated through a case study from a recent 4-D survey over the Gullfaks field. 2 figs.

Seismic Fragility Analysis of a Condensate Storage Tank with Age-Related Degradations

Energy Technology Data Exchange (ETDEWEB)

Nie, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Braverman, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hofmayer, C [Brookhaven National Lab. (BNL), Upton, NY (United States); Choun, Y-S [Brookhaven National Lab. (BNL), Upton, NY (United States); Kim, MK [Brookhaven National Lab. (BNL), Upton, NY (United States); Choi, I-K [Brookhaven National Lab. (BNL), Upton, NY (United States)

2011-04-01

The Korea Atomic Energy Research Institute (KAERI) is conducting a five-year research project to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). The KAERI research project includes three specific areas that are essential to seismic probabilistic risk assessment (PRA): (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. Since 2007, Brookhaven National Laboratory (BNL) has entered into a collaboration agreement with KAERI to support its development of seismic capability evaluation technology for degraded structures and components. The collaborative research effort is intended to continue over a five year period. The goal of this collaboration endeavor is to assist KAERI to develop seismic fragility analysis methods that consider the potential effects of age-related degradation of structures, systems, and components (SSCs). The research results of this multi-year collaboration will be utilized as input to seismic PRAs. This report describes the research effort performed by BNL for the Year 4 scope of work. This report was developed as an update to the Year 3 report by incorporating a major supplement to the Year 3 fragility analysis. In the Year 4 research scope, an additional study was carried out to consider an additional degradation scenario, in which the three basic degradation scenarios, i.e., degraded tank shell, degraded anchor bolts, and cracked anchorage concrete, are combined in a non-perfect correlation manner. A representative operational water level is used for this effort. Building on the same CDFM procedure implemented for the Year 3 Tasks, a simulation method was applied using optimum Latin Hypercube samples to characterize the deterioration behavior of the fragility capacity as a function of age-related degradations. The results are summarized in Section 5

Seismic re-evaluation of piping systems of heavy water plant, Kota

CERN Document Server

Mishra, R; Soni, R S; Venkat-Raj, V

2002-01-01

Heavy Water Plant, Kota is the first indigenous heavy water plant built in India. The plant started operation in the year 1985 and it is approaching the completion of its originally stipulated design life. In view of the excellent record of plant operation for the past so many years, it has been planned to carry out various exercises for the life extension of the plant. In the first stage, evaluation of operation stresses was carried out for the process critical piping layouts and equipment, which are connected with 25 process critical nozzle locations, identified based on past history of the plant performance. Fatigue life evaluation has been carried out to fmd out the Cumulative Usage Factor, which helps in arriving at a decision regarding the life extension of the plant. The results of these exercises have been already reported separately vide BARC/200I /E/O04. In the second stage, seismic reevaluation of the plant has been carried out to assess its ability to maintain its integ:rity in case of a seismic e...

Seismic microzonation of Bangalore, India

Indian Academy of Sciences (India)

Evaluation of seismic hazards and microzonation of cities enable us to characterize the potential seismic areas which have similar exposures to haz- ards of earthquakes, and these results can be used for designing new structures or retrofitting the existing ones. Study of seismic hazard and preparation of microzonation ...

Simple procedure for evaluating earthquake response spectra of large-event motions based on site amplification factors derived from smaller-event records

International Nuclear Information System (INIS)

Dan, Kazuo; Miyakoshi, Jun-ichi; Yashiro, Kazuhiko.

1996-01-01

A primitive procedure was proposed for evaluating earthquake response spectra of large-event motions to make use of records from smaller events. The result of the regression analysis of the response spectra was utilized to obtain the site amplification factors in the proposed procedure, and the formulation of the seismic-source term in the regression analysis was examined. A linear form of the moment magnitude, Mw, is good for scaling the source term of moderate earthquakes with Mw of 5.5 to 7.0, while a quadratic form of Mw and the ω-square source-spectrum model is appropriate for scaling the source term of smaller and greater earthquakes, respectively. (author). 52 refs

Korea-Japan Joint Research on Development of Seismic Capacity Evaluation and Enhancement Technology Considering Near-Fault Effect (Final Report)

Energy Technology Data Exchange (ETDEWEB)

Choun, Young Sun; Choi, In Kil; Kim, Min Kyu [KAERI, Daejeon (Korea, Republic of); Ohtori, Yasuki; Shiba, Yoshiaki; Nakajima, Masato [Central Research Institute of Electric Power Industry, Tokyo (Japan)

2006-12-15

We compiled the results of the source analysis obtained under the collaboration research. Recent construction scheme for source modeling adopted in Japan is described, and strong-motion prediction is performed assuming the scenario earthquakes occurring in the Ulsan fault system, Korea. Finally Qs values beneath the Korean inland crust are estimated using strong-motion records in Korea from the 2005 Off West Fukuoka earthquake (M7.0). Probabilistic seismic hazard for four NPP sites in Korea are evaluated, in which the site specific attenuation equations with Index SA developed for NPP sites are adopted. Furthermore, the uniform hazard spectra for the four NPP sites in Korea are obtained by conducting the PSHA by using the attenuation equations with the index of response spectra and seismic source model cases with maximum weights. The supporting tools for seismic response analysis, the evaluation tool for evaluating annual probability of failure, and system analysis program were developed for the collaboration. The tools were verified with theoretical results, the results written in the reference document of EQESRA, and so forth. The system analysis program was applied for the investigation of the effect of improving the seismic capacity of equipment. We evaluated the annual probability of failure of isolated and non-isolated EDG at Younggwang NPP site as the results of the collaboration. The input ground motion for generating the seismic fragility curve was determined based on the seismic hazard analysis. It was found that the annual probability of failure of isolated EDG is lower than that of non-isolated EDG.

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

Nuclear power plant of Fessenheim: evaluation of the seismic risk; Centrale Nucleaire de Fessenheim: appreciation du risque sismique

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

The seismic risk taken into account during the sizing of the nuclear power plant of Fessenheim seems to have been under evaluated at this time. The revaluation of the seismic risk, as proposed, until this day by EDF in order to the third ten-year visit of the power plant, planned for 2009, leads to a significant under evaluation of the risk and then is not acceptable. The present expertise details point by point the weaknesses of these revaluation. The power plant has been sized in an elastic manner that is generally strongly for the safety side. It is imperative to proceed the most quickly as possible to a deep control of the seismic resistance of the power plant of Fessenheim and then after having proceeded to a revision of the seismic risk in taking into account the actual knowledge in this field. (N.C.)

CARES (Computer Analysis for Rapid Evaluation of Structures) Version 1.0, seismic module

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulas, A.J.; Miller, C.A.; Costantino, C.J.

1990-07-01

During FY's 1988 and 1989, Brookhaven National Laboratory (BNL) developed the CARES system (Computer Analysis for Rapid Evaluation of Structures) for the US Nuclear Regulatory Commission (NRC). CARES is a PC software system which has been designed to perform structural response computations similar to those encountered in licensing reviews of nuclear power plant structures. The documentation of the Seismic Module of CARES consists of three volumes. This report represents Volume 3 of the volume documentation of the Seismic Module of CARES. It presents three sample problems typically encountered in the Soil-Structure Interaction analyses. 14 refs., 36 figs., 2 tabs

EVALUATION ON THE SEISMIC RESPONSE CHARACTERISTICS OF A ROAD EMBANKMENT BASED ON THE MODERATE EARTHQUAKE OBSERVATION AND THE MICROTREMOR MEASUREMENT

Science.gov (United States)

Hata, Yoshiya; Ichii, Koji; Yamada, Masayuki; Tokida, Ken-Ichi; Takezawa, Koichiro; Shibao, Susumu; Mitsushita, Junji; Murata, Akira; Furukawa, Aiko; Koizumi, Keigo

Accurate evaluation on the seismic response characteristics of a road embankment is very important for the rational seismic assessment. However, in a lot of previous studies, the seismic response characteristics of an embankment were evaluated based on the results of shaking table test, centrifuge model test and dynamic FEM analysis. In this study, the transfer function and the shear wave velocity of a road embankment were evaluated based on the in-situ records of moderate earthquake observation and microtremor measurement. Test results show the possibility that the shear wave velocity of an embankment can be estimated by the earthquake observation or the microtremor measurement and the dynamic linear FEM analysis.

Seismic evaluation of critical facilities at the Lawrence Livermore Laboratory

International Nuclear Information System (INIS)

Murray, R.C.; Tokarz, F.J.

1976-01-01

The performance of critical facilities at the Lawrence Livermore Laboratory (LLL) are being evaluated for severe earthquake loading. Facilities at Livermore, Site-300 and the Nevada Test Site are included in this study. These facilities are identified, the seismic criteria used for the analysis are indicated, the various methods used for structural analysis are discussed and a summary of the results of facilities analyzed to date are presented

Seismic evaluation of nuclear installations; Avaliacao sismica de instalacoes nucleares

Energy Technology Data Exchange (ETDEWEB)

Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

1997-10-01

Some considerations regarding extreme external events, natural or man-induce, such as earthquakes, floods, air crashes, etc, shall be done for nuclear facilities to minimizing the potential impact of the installation on the public and the environment. In this paper the main aspects of the seismic evaluation of nuclear facilities (except the nuclear power reactors) will be presented based on different codes and standards. (author). 7 refs., 2 tabs.

A report on seismic re-evaluation of Cirus systems

International Nuclear Information System (INIS)

Varma, Veto; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

2003-06-01

Cirus was initiated way back in 1955 and its design was made with the methods prevailing at that time. The design codes and safety standards have changed since then, particularly with respect to seismic design criteria. As the structure is an important safety related structure it is mandatory to meet the present statutory requirement. This report contains the seismic qualification for some of the Cirus systems. The report has four parts. Part I gives the analytical studies performed in the containment building, Part II describes of experimental studies carried out to validate the analytical studies for containment builaing, Part III explains the seismic retrofitting of Battery bank, and Part IV summarizes the seismic qualification of inlet and exhaust damper of Cirus. (author)

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)

Seismic Performance Evaluation of Concrete Gravity Dams with Penetrated Cracks Considering Fluid–Structure Interaction

Directory of Open Access Journals (Sweden)

A. Behshad

2018-02-01

Full Text Available In this paper, a comprehensive study on the seismic behavior of fractured concrete gravity dams during ground shakings is carried out considering dam–reservoir interaction effects. To gain the seismic behavior of the whole system, finite and boundary elements are employed to model the liquid region and the cracked structure, respectively. Formulation and different computational aspects of the suggested staggered hybrid approach are thoroughly argued. A computer code was developed in order to discuss the presented hybrid BE–DE technique and comparisons are made between the obtained results and those reported in the literature. To gain this goal, several problems of seismic excitations in frequency- and time-domains are presented employing the proposed approach, showing that the present results agree well with the results from other numerical procedures. The cracked Koyna Dam is scrutinized, considering the dynamic interaction between dam and reservoir with focus on the nonlinear behavior due to its top profile crack. The developed numerical model is rigorously validated by extensive comparisons with available results in the literature in which the dam–reservoir interaction were simplified by added masses. It can be concluded that there is significant disparity between the overturning and sliding response schemes of the nonlinear analysis and those of added mass technique.

Evaluation of potential surface rupture and review of current seismic hazards program at the Los Alamos National Laboratory. Final report

International Nuclear Information System (INIS)

1991-01-01

This report summarizes the authors review and evaluation of the existing seismic hazards program at Los Alamos National Laboratory (LANL). The report recommends that the original program be augmented with a probabilistic analysis of seismic hazards involving assignment of weighted probabilities of occurrence to all potential sources. This approach yields a more realistic evaluation of the likelihood of large earthquake occurrence particularly in regions where seismic sources may have recurrent intervals of several thousand years or more. The report reviews the locations and geomorphic expressions of identified fault lines along with the known displacements of these faults and last know occurrence of seismic activity. Faults are mapped and categorized into by their potential for actual movement. Based on geologic site characterization, recommendations are made for increased seismic monitoring; age-dating studies of faults and geomorphic features; increased use of remote sensing and aerial photography for surface mapping of faults; the development of a landslide susceptibility map; and to develop seismic design standards for all existing and proposed facilities at LANL

Seismic component fragility data base for IPEEE

International Nuclear Information System (INIS)

Bandyopadhyay, K.; Hofmayer, C.

1990-01-01

Seismic probabilistic risk assessment or a seismic margin study will require a reliable data base of seismic fragility of various equipment classes. Brookhaven National Laboratory (BNL) has selected a group of equipment and generically evaluated the seismic fragility of each equipment class by use of existing test data. This paper briefly discusses the evaluation methodology and the fragility results. The fragility analysis results when used in the Individual Plant Examination for External Events (IPEEE) Program for nuclear power plants are expected to provide insights into seismic vulnerabilities of equipment for earthquakes beyond the design basis. 3 refs., 1 fig., 1 tab

Seismic Safety Margins Research Program: Phase II program plan (FY 83-FY 84)

International Nuclear Information System (INIS)

Bohn, M.P.; Bernreuter, D.L.; Cover, L.E.; Johnson, J.J.; Shieh, L.C.; Shukla, S.N.; Wells, J.E.

1982-01-01

The Seismic Safety Margins Research Program (SSMRP) is an NRC-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its goal is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from a commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis and design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. As currently planned, the SSMRP will be completed in September, 1984. This document presents the program plan for work to be done during the remainder of the program. In Phase I of the SSMRP, the necessary tools (both computer codes and data bases) for performing a detailed seismic risk analysis were identified and developed. Demonstration calculations were performed on the Zion Nuclear Power Plant. In the remainder of the program (Phase II) work will be concentrated on developing a simplified SSMRP methodology for routine probabilistic risk assessments, quantitative validation of the tools developed and application of the simplified methodology to a Boiling Water Reactor. (The Zion plant is a pressurized water reactor.) In addition, considerable effort will be devoted to making the codes and data bases easily accessible to the public

Seismic rupture modelling, strong motion prediction and seismic hazard assessment: fundamental and applied approaches; Modelisation de la rupture sismique, prediction du mouvement fort, et evaluation de l'alea sismique: approches fondamentale et appliquee

Energy Technology Data Exchange (ETDEWEB)

Berge-Thierry, C

2007-05-15

The defence to obtain the 'Habilitation a Diriger des Recherches' is a synthesis of the research work performed since the end of my Ph D. thesis in 1997. This synthesis covers the two years as post doctoral researcher at the Bureau d'Evaluation des Risques Sismiques at the Institut de Protection (BERSSIN), and the seven consecutive years as seismologist and head of the BERSSIN team. This work and the research project are presented in the framework of the seismic risk topic, and particularly with respect to the seismic hazard assessment. Seismic risk combines seismic hazard and vulnerability. Vulnerability combines the strength of building structures and the human and economical consequences in case of structural failure. Seismic hazard is usually defined in terms of plausible seismic motion (soil acceleration or velocity) in a site for a given time period. Either for the regulatory context or the structural specificity (conventional structure or high risk construction), seismic hazard assessment needs: to identify and locate the seismic sources (zones or faults), to characterize their activity, to evaluate the seismic motion to which the structure has to resist (including the site effects). I specialized in the field of numerical strong-motion prediction using high frequency seismic sources modelling and forming part of the IRSN allowed me to rapidly working on the different tasks of seismic hazard assessment. Thanks to the expertise practice and the participation to the regulation evolution (nuclear power plants, conventional and chemical structures), I have been able to work on empirical strong-motion prediction, including site effects. Specific questions related to the interface between seismologists and structural engineers are also presented, especially the quantification of uncertainties. This is part of the research work initiated to improve the selection of the input ground motion in designing or verifying the stability of structures. (author)

Linearized inversion frameworks toward high-resolution seismic imaging

KAUST Repository

Aldawood, Ali

2016-09-01

internally multiply scattered seismic waves to obtain highly resolved images delineating vertical faults that are otherwise not easily imaged by primaries. Seismic interferometry is conventionally based on the cross-correlation and convolution of seismic traces to transform seismic data from one acquisition geometry to another. The conventional interferometric transformation yields virtual data that suffers from low temporal resolution, wavelet distortion, and correlation/convolution artifacts. I therefore incorporate a least-squares datuming technique to interferometrically transform vertical-seismic-profile surface-related multiples to surface-seismic-profile primaries. This yields redatumed data with high temporal resolution and less artifacts, which are subsequently imaged to obtain highly resolved subsurface images. Tests on synthetic examples demonstrate the efficiency of the proposed techniques, yielding highly resolved migrated sections compared with images obtained by imaging conventionally redatumed data. I further advance the recently developed cost-effective Generalized Interferometric Multiple Imaging procedure, which aims to not only image first but also higher-order multiples as well. I formulate this procedure as a linearized inversion framework and solve it as a least-squares problem. Tests of the least-squares Generalized Interferometric Multiple imaging framework on synthetic datasets and demonstrate that it could provide highly resolved migrated images and delineate vertical fault planes compared with the standard procedure. The results support the assertion that this linearized inversion framework can illuminate subsurface zones that are mainly illuminated by internally scattered energy.

PSHA in Israel by using the synthetic ground motions from simulated seismicity: the modified SvE procedure

Science.gov (United States)

Meirova, T.; Shapira, A.; Eppelbaum, L.

2018-05-01

In this study, we updated and modified the SvE approach of Shapira and van Eck (Nat Hazards 8:201-215, 1993) which may be applied as an alternative to the conventional probabilistic seismic hazard assessment (PSHA) in Israel and other regions of low and moderate seismicity where measurements of strong ground motions are scarce. The new computational code SvE overcomes difficulties associated with the description of the earthquake source model and regional ground-motion scaling. In the modified SvE procedure, generating suites of regional ground motion is based on the extended two-dimensional source model of Motazedian and Atkinson (Bull Seism Soc Amer 95:995-1010, 2005a) and updated regional ground-motion scaling (Meirova and Hofstteter, Bull Earth Eng 15:3417-3436, 2017). The analytical approach of Mavroeidis and Papageorgiou (Bull Seism Soc Amer 93:1099-1131, 2003) is used to simulate the near-fault acceleration with the near-fault effects. The comparison of hazard estimates obtained by using the conventional method implemented in the National Building Code for Design provisions for earthquake resistance of structures and the modified SvE procedure for rock-site conditions indicates a general agreement with some perceptible differences at the periods of 0.2 and 0.5 s. For the periods above 0.5 s, the SvE estimates are systematically greater and can increase by a factor of 1.6. For the soft-soil sites, the SvE hazard estimates at the period of 0.2 s are greater than those based on the CB2008 ground-motion prediction equation (GMPE) by a factor of 1.3-1.6. We suggest that the hazard estimates for the sites with soft-soil conditions calculated by the modified SvE procedure are more reliable than those which can be found by means of the conventional PSHA. This result agrees with the opinion that the use of a standard GMPE applying the NEHRP soil classification based on the V s, 30 parameter may be inappropriate for PSHA at many sites in Israel.

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

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

Evaluation of the seismic integrity of a plutonium-handling facility

International Nuclear Information System (INIS)

Coats, D.W.

1981-01-01

Many studies have been made by and for the Lawrence Livermore National Laboratory (LLNL) to ensure the seismic safety of its Plutonium Facility (Building 332). These studies have included seismological and geologic field investigations to define the actual seismic hazard existing at the Laboratory site as well as structural studies of the Facility itself. Because the basic seismic design criteria has undergone changes over the years, numerous structural studies and upgrades have been completed. The seismic criteria in use at the LLNL site is reviewed on a continuing basis as new information on the seismicity and geology of the Livermore Valley is obtained. At present, the Laboratory's Earth Sciences Division is conducting a multi-million dollar program to identify and characterize the geologic hazards at the Livermore site, with the primary emphasis on earthquake hazards in the Livermore Valley. This effort is undergoing an independent review by Woodward-Clyde Associates. Additionally, because of increased concerns over the seismic safety of Building 332, the Laboratory has initiated an independent structural review. This review effort will be monitored by the California Seismic Safety Commission to ensure its independence. Both of these studiies are in their early stages and results are not yet available

Seismic rupture modelling, strong motion prediction and seismic hazard assessment: fundamental and applied approaches

International Nuclear Information System (INIS)

Berge-Thierry, C.

2007-05-01

The defence to obtain the 'Habilitation a Diriger des Recherches' is a synthesis of the research work performed since the end of my Ph D. thesis in 1997. This synthesis covers the two years as post doctoral researcher at the Bureau d'Evaluation des Risques Sismiques at the Institut de Protection (BERSSIN), and the seven consecutive years as seismologist and head of the BERSSIN team. This work and the research project are presented in the framework of the seismic risk topic, and particularly with respect to the seismic hazard assessment. Seismic risk combines seismic hazard and vulnerability. Vulnerability combines the strength of building structures and the human and economical consequences in case of structural failure. Seismic hazard is usually defined in terms of plausible seismic motion (soil acceleration or velocity) in a site for a given time period. Either for the regulatory context or the structural specificity (conventional structure or high risk construction), seismic hazard assessment needs: to identify and locate the seismic sources (zones or faults), to characterize their activity, to evaluate the seismic motion to which the structure has to resist (including the site effects). I specialized in the field of numerical strong-motion prediction using high frequency seismic sources modelling and forming part of the IRSN allowed me to rapidly working on the different tasks of seismic hazard assessment. Thanks to the expertise practice and the participation to the regulation evolution (nuclear power plants, conventional and chemical structures), I have been able to work on empirical strong-motion prediction, including site effects. Specific questions related to the interface between seismologists and structural engineers are also presented, especially the quantification of uncertainties. This is part of the research work initiated to improve the selection of the input ground motion in designing or verifying the stability of structures. (author)

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

CARES (Computer Analysis for Rapid Evaluation of Structures) Version 1.0, seismic module

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulas, A.J.; Miller, C.A.; Costantino, C.J.

1990-07-01

During FY's 1988 and 1989, Brookhaven National Laboratory (BNL) developed the CARES system (Computer Analysis for Rapid Evaluation of Structures) for the US Nuclear Regulatory Commission (NRC). CARES is a PC software system which has been designed to perform structural response computations similar to those encountered in licencing reviews of nuclear power plant structures. The docomentation of the Seismic Module of CARES consists of three volumes. This report represents Volume 1 of the three volume documentation of the Seismic Module of CARES. It concentrates on the theoretical basis of the system and presents modeling assumptions and limitations as well as solution schemes and algorithms of CARES. 31 refs., 6 figs

Seismic PSA of nuclear power plants a case study

International Nuclear Information System (INIS)

Hari Prasad, M.; Dubey, P.N.; Reddy, G.R.; Saraf, R.K.; Ghosh, A.K.

2006-07-01

Seismic Probabilistic Safety Assessment (Seismic PSA) analysis is an external event PSA analysis. The objective of seismic PSA for the plants is to examine the existence of plant vulnerabilities against postulated earthquakes by numerically assessing the plant safety and to take appropriate measures to enhance the plant safety. Seismic PSA analysis integrates the seismic hazard analysis, seismic response analysis, seismic fragility analysis and system reliability/ accident sequence analysis. In general, the plant consists of normally operating and emergency standby systems and components. The failure during an earthquake (induced directly by excessive inertial stresses or indirectly following the failure of some other item) of an operating component will lead to a change in the state of the plant. In that case, various scenarios can follow depending on the initiating event and the status of other sub-systems. The analysis represents these possible chronological sequences by an event tree. The event trees and the associated fault trees model the sub-systems down to the level of individual components. The procedure has been applied for a typical Indian nuclear power plant. From the internal event PSA level I analysis significant contribution to the Core Damage Frequency (CDF) was found due to the Fire Water System. Hence, this system was selected to establish the procedure of seismic PSA. In this report the different elements that go into seismic PSA analysis have been discussed. Hazard curves have been developed for the site. Fragility curve for the seismically induced failure of Class IV power has been developed. The fragility curve for fire-water piping system has been generated. Event tree for Class IV power supply has been developed and the dominating accident sequences were identified. CDF has been estimated from these dominating accident sequences by convoluting hazard curves of initiating event and fragility curves of the safety systems. (author)

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1993-01-01

An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

Seismic verification methods for structures and equipment of VVER-type and RBMK-type NPPs (summary of experiences)

International Nuclear Information System (INIS)

Masopust, R.

2003-01-01

The main verification methods for structures and equipment of already existing VVER-type and RBMK-type NPPs are briefly described. The following aspects are discussed: fundamental seismic safety assessment principles for VVER/RBMK-type NPPs (seismic safety assessment procedure, typical work plan for seismic safety assessment of existing NPPs, SMA (HCLPF) calculations, modified GIP (GIP-VVER) procedure, similarity of VVER/RBMK equipment to that included in the SQUG databases and seismic interactions

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)

Seismic rupture modelling, strong motion prediction and seismic hazard assessment: fundamental and applied approaches; Modelisation de la rupture sismique, prediction du mouvement fort, et evaluation de l'alea sismique: approches fondamentale et appliquee

Energy Technology Data Exchange (ETDEWEB)

Berge-Thierry, C

2007-05-15

The defence to obtain the 'Habilitation a Diriger des Recherches' is a synthesis of the research work performed since the end of my Ph D. thesis in 1997. This synthesis covers the two years as post doctoral researcher at the Bureau d'Evaluation des Risques Sismiques at the Institut de Protection (BERSSIN), and the seven consecutive years as seismologist and head of the BERSSIN team. This work and the research project are presented in the framework of the seismic risk topic, and particularly with respect to the seismic hazard assessment. Seismic risk combines seismic hazard and vulnerability. Vulnerability combines the strength of building structures and the human and economical consequences in case of structural failure. Seismic hazard is usually defined in terms of plausible seismic motion (soil acceleration or velocity) in a site for a given time period. Either for the regulatory context or the structural specificity (conventional structure or high risk construction), seismic hazard assessment needs: to identify and locate the seismic sources (zones or faults), to characterize their activity, to evaluate the seismic motion to which the structure has to resist (including the site effects). I specialized in the field of numerical strong-motion prediction using high frequency seismic sources modelling and forming part of the IRSN allowed me to rapidly working on the different tasks of seismic hazard assessment. Thanks to the expertise practice and the participation to the regulation evolution (nuclear power plants, conventional and chemical structures), I have been able to work on empirical strong-motion prediction, including site effects. Specific questions related to the interface between seismologists and structural engineers are also presented, especially the quantification of uncertainties. This is part of the research work initiated to improve the selection of the input ground motion in designing or verifying the stability of structures. (author)

New seismic attributes and methodology for automated stratigraphic, structural, and reservoir analysis

Energy Technology Data Exchange (ETDEWEB)

Randen, Trygve; Reymond, Benoit; Sjulstad, Hans Ivar; Soenneland, Lars

1998-12-31

Seismic stratigraphy represents an attractive framework for interpretation of 3-D data. This presentation is an introduction to a set of primitives that will enable guided interpretation of seismic signals in the framework of seismic stratigraphy. A method capable of automatic detection of terminations is proposed. The new procedure can be run on the entire seismic volume or it may be restricted to a limited time interval and detects terminations in an unguided manner without prior interpretation. The density of terminations can be computed. The procedure may alternatively be guided by pre-existing interpretation, e.g. detecting terminations onto an interpreted horizon. In such a case, the density of terminations will be a new surface attribute. 6 refs., 3 figs.

New seismic attributes and methodology for automated stratigraphic, structural, and reservoir analysis

Energy Technology Data Exchange (ETDEWEB)

Randen, Trygve; Reymond, Benoit; Sjulstad, Hans Ivar; Soenneland, Lars

1999-12-31

Seismic stratigraphy represents an attractive framework for interpretation of 3-D data. This presentation is an introduction to a set of primitives that will enable guided interpretation of seismic signals in the framework of seismic stratigraphy. A method capable of automatic detection of terminations is proposed. The new procedure can be run on the entire seismic volume or it may be restricted to a limited time interval and detects terminations in an unguided manner without prior interpretation. The density of terminations can be computed. The procedure may alternatively be guided by pre-existing interpretation, e.g. detecting terminations onto an interpreted horizon. In such a case, the density of terminations will be a new surface attribute. 6 refs., 3 figs.

Seismic behaviour of geotechnical structures

Directory of Open Access Journals (Sweden)

F. Vinale

2002-06-01

Full Text Available This paper deals with some fundamental considerations regarding the behaviour of geotechnical structures under seismic loading. First a complete definition of the earthquake disaster risk is provided, followed by the importance of performing site-specific hazard analysis. Then some suggestions are provided in regard to adequate assessment of soil parameters, a crucial point to properly analyze the seismic behaviour of geotechnical structures. The core of the paper is centered on a critical review of the analysis methods available for studying geotechnical structures under seismic loadings. All of the available methods can be classified into three main classes, including the pseudo-static, pseudo-dynamic and dynamic approaches, each of which is reviewed for applicability. A more advanced analysis procedure, suitable for a so-called performance-based design approach, is also described in the paper. Finally, the seismic behaviour of the El Infiernillo Dam was investigated. It was shown that coupled elastoplastic dynamic analyses disclose some of the important features of dam behaviour under seismic loading, confirmed by comparing analytical computation and experimental measurements on the dam body during and after a past earthquake.

Integral anomalous effect of an oil and gas deposit in a seismic wave field

Energy Technology Data Exchange (ETDEWEB)

Korostyshevskiy, M.B.; Nabokov, G.N.

1981-01-01

The basic precepts of an elaborated version of a procedure for forecasting (direct exploration) of oil and gas deposits according to seismic prospecting data MOV are examined. This procedure was previously called the procedure of analysis of the integral affect of an oil and gas deposit in a seismic wave field (MIIEZ-VP). The procedure is implemented in the form of an automated system ASOM-VP for the BESM-4 computer in a standard configuration equipped with standard input-output devices for seismic information (''Potok'', MVU, ''Atlas''). The entire procedure of processing from input of data into the computer to output of resulting maps and graphs on graph plotter ''Atlas'' is automated. Results of testing of procedure MIIEZ-VP and system ASOM-VP on drilled areas of Kazakhstan, Azerbaydzhan and Uzbekistan are cited.

Evaluating Seismic Activity in Ethiopia

African Journals Online (AJOL)

map is constructed from which seismic risks in a given sector ... troyed (10, 11) and the people of Eritrea remember these years ... terms of damage caused to man-made structures; they refer to .... walls of a well designed modern building were deta- ched from ... Although, at present, no theory is satisfactory, the fact remains.

Integral Analysis of Seismic Refraction and Ambient Vibration Survey for Subsurface Profile Evaluation

Science.gov (United States)

Hazreek, Z. A. M.; Kamarudin, A. F.; Rosli, S.; Fauziah, A.; Akmal, M. A. K.; Aziman, M.; Azhar, A. T. S.; Ashraf, M. I. M.; Shaylinda, M. Z. N.; Rais, Y.; Ishak, M. F.; Alel, M. N. A.

2018-04-01

Geotechnical site investigation as known as subsurface profile evaluation is the process of subsurface layer characteristics determination which finally used for design and construction phase. Traditionally, site investigation was performed using drilling technique thus suffers from several limitation due to cost, time, data coverage and sustainability. In order to overcome those problems, this study adopted surface techniques using seismic refraction and ambient vibration method for subsurface profile depth evaluation. Seismic refraction data acquisition and processing was performed using ABEM Terraloc and OPTIM software respectively. Meanwhile ambient vibration data acquisition and processing was performed using CityShark II, Lennartz and GEOPSY software respectively. It was found that studied area consist of two layers representing overburden and bedrock geomaterials based on p-wave velocity value (vp = 300 – 2500 m/s and vp > 2500 m/s) and natural frequency value (Fo = 3.37 – 3.90 Hz) analyzed. Further analysis found that both methods show some good similarity in term of depth and thickness with percentage accuracy at 60 – 97%. Consequently, this study has demonstrated that the application of seismic refractin and ambient vibration method was applicable in subsurface profile depth and thickness estimation. Moreover, surface technique which consider as non-destructive method adopted in this study was able to compliment conventional drilling method in term of cost, time, data coverage and environmental sustainaibility.

Evaluation of structural fragilities for an IPEEE seismic probabilistic risk assessment study

International Nuclear Information System (INIS)

Ghiocel, D.M.; Wilson, P.R.; Stevenson, J.D.

1995-01-01

The paper presents the main issues and results of a structural fragility analysis for a Seismic Probabilistic Risk Assessment (SPRA) study of a nuclear power plant (NPP) in the Eastern US. The fragility evaluations were performed for the Reactor Building, Auxiliary Building, Intake Structure and Diesel Generator Building. The random seismic input is defined in terms of the Uniform Hazard Spectrum (UHS) earthquake on the NPP site anchored to a reference level of 0.40 g Zero Period Ground Acceleration (ZPGA). Because of the soft soil conditions new Soil-Structure Interaction (SSI) analyses were performed using the original finite element (stick) structural models and the complex frequency approach. The soil deposit randomness was described by the variations in both the low strain soil shear modules and in its dependence with the shear strain. The probabilistic SSI analyses were performed using digital simulation techniques. The critical failure modes for each structure are investigated and the fragility evaluations are discussed. Concluding remarks and recommendations for improving the quality of the structural fragility analyses are included

GUI program to compute probabilistic seismic hazard analysis

International Nuclear Information System (INIS)

Shin, Jin Soo; Chi, H. C.; Cho, J. C.; Park, J. H.; Kim, K. G.; Im, I. S.

2006-12-01

The development of program to compute probabilistic seismic hazard is completed based on Graphic User Interface(GUI). The main program consists of three part - the data input processes, probabilistic seismic hazard analysis and result output processes. The probabilistic seismic hazard analysis needs various input data which represent attenuation formulae, seismic zoning map, and earthquake event catalog. The input procedure of previous programs based on text interface take a much time to prepare the data. The data cannot be checked directly on screen to prevent input erroneously in existing methods. The new program simplifies the input process and enable to check the data graphically in order to minimize the artificial error within limits of the possibility

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

LANL seismic screening method for existing buildings

International Nuclear Information System (INIS)

Dickson, S.L.; Feller, K.C.; Fritz de la Orta, G.O.

1997-01-01

The purpose of the Los Alamos National Laboratory (LANL) Seismic Screening Method is to provide a comprehensive, rational, and inexpensive method for evaluating the relative seismic integrity of a large building inventory using substantial life-safety as the minimum goal. The substantial life-safety goal is deemed to be satisfied if the extent of structural damage or nonstructural component damage does not pose a significant risk to human life. The screening is limited to Performance Category (PC) -0, -1, and -2 buildings and structures. Because of their higher performance objectives, PC-3 and PC-4 buildings automatically fail the LANL Seismic Screening Method and will be subject to a more detailed seismic analysis. The Laboratory has also designated that PC-0, PC-1, and PC-2 unreinforced masonry bearing wall and masonry infill shear wall buildings fail the LANL Seismic Screening Method because of their historically poor seismic performance or complex behavior. These building types are also recommended for a more detailed seismic analysis. The results of the LANL Seismic Screening Method are expressed in terms of separate scores for potential configuration or physical hazards (Phase One) and calculated capacity/demand ratios (Phase Two). This two-phase method allows the user to quickly identify buildings that have adequate seismic characteristics and structural capacity and screen them out from further evaluation. The resulting scores also provide a ranking of those buildings found to be inadequate. Thus, buildings not passing the screening can be rationally prioritized for further evaluation. For the purpose of complying with Executive Order 12941, the buildings failing the LANL Seismic Screening Method are deemed to have seismic deficiencies, and cost estimates for mitigation must be prepared. Mitigation techniques and cost-estimate guidelines are not included in the LANL Seismic Screening Method

Survey of seismic fragilities used in PRA studies of nuclear power plants

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.; Chokshi, N.C.

1998-01-01

In recent years, seismic PRA studies have been performed on a large number of nuclear power plants in the USA. This paper presents a summary of a survey on fragility databases and the range of evaluated fragility values of various equipment categories based on past PRAs. The survey includes the use of experience data, the interpretations of available test data, and the quantification of uncertainties. The surveyed fragility databases are limited to data available in the public domain such as NUREG reports, conference proceedings and other publicly available reports. The extent of the availability of data as well as limitations are studied and tabulated for various equipment categories. The survey of the fragility values in past PRA studies includes not only the best estimate values, but also the dominant failure modes and the estimated uncertainty levels for each equipment category. The engineering judgments employed in estimating the uncertainty in the fragility values are also studied. This paper provides a perspective on the seismic fragility evaluation procedures for equipment in order to clearly identify the engineering analysis and judgment used in past seismic PRA studies

Component evaluation testing and analysis algorithms.

Energy Technology Data Exchange (ETDEWEB)

Hart, Darren M.; Merchant, Bion John

2011-10-01

The Ground-Based Monitoring R&E Component Evaluation project performs testing on the hardware components that make up Seismic and Infrasound monitoring systems. The majority of the testing is focused on the Digital Waveform Recorder (DWR), Seismic Sensor, and Infrasound Sensor. In order to guarantee consistency, traceability, and visibility into the results of the testing process, it is necessary to document the test and analysis procedures that are in place. Other reports document the testing procedures that are in place (Kromer, 2007). This document serves to provide a comprehensive overview of the analysis and the algorithms that are applied to the Component Evaluation testing. A brief summary of each test is included to provide the context for the analysis that is to be performed.

A development of an evaluation flow chart for seismic stability of rock slopes based on relations between safety factor and sliding failure

International Nuclear Information System (INIS)

Kawai, Tadashi; Ishimaru, Makoto

2010-01-01

Recently, it is necessary to assess quantitatively seismic safety of critical facilities against the earthquake- induced rock slope failure from the viewpoint of seismic PSA. Under these circumstances, it is needed to evaluate the seismic stability of surrounding slopes against extremely strong ground motions. In order to evaluate the seismic stability of surrounding slopes, the most conventional method is to compare safety factors on an expected sliding surface, which is calculated from the stability analysis based on the limit equilibrium concept, to a critical value which judges stability or instability. The method is very effective to examine whether or not the sliding surface is safe. However, it does not mean that the sliding surface falls whenever the safety factor becomes smaller than the critical value during an earthquake. Therefore the authors develop a new evaluation flow chart for the seismic stability of rock slopes based on relations between safety factor and sliding failure. Furthermore, the developed flow chart was validated by comparing two kinds of safety factors calculated from a centrifuge test result concerned with a rock slope. (author)

The guideline and practical procedures for earthquake-resistant design of nuclear power plants in Japan

International Nuclear Information System (INIS)

Watabe, M.

1985-01-01

The Guideline for the aseismic design of nuclear reactor facilities, revised in 1981, is introduced. The basic philosophy entails structural integrity against a major earthquake, rigid structure for less deformation and foundation on rock. The classification of facilities is then explained. Some practical examples are tabulated. In the light of the above classifications, evaluation procedures for aseismic design are defined. Design basis earthquake ground motions, S1 and S2, are defined. S1 is the maximum possible earthquake ground motion, while S2 is the maximum credible one. The relation between active faults and S1, S2 motions is explained, seismic forces induced by S1 and S2 are expressed in terms of response spectra. Static seismic coefficient procedures are also applied to evaluate seismic forces, as a minimum guide-line based on dynamic analysis. Combinations of seismic forces and allowable limits are then explained. In the second part of the paper, seismic analysis for reactor buildings as a part of design practice is outlined. There are three major key points in practical aseismic design. The first one is input design earthquake motions, in which soil/foundation interaction problems are also included. In practice, ground motions at the free field rock surface have to be convoluted or deconvoluted to obtain base rock motions, which are applied to estimate input design earthquake motions by way of finite element analysis or a lumped mass lattice model. Also introduced is dynamic modelling of the reactor building with its non-linear behaviour represented by plastic deformation of reinforced concrete members as well as by uplift characteristics of foundations. Then an evaluation of aseismic safety is introduced. (author)

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

Energy Technology Data Exchange (ETDEWEB)

E.N. Lindner

2004-12-03

The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

International Nuclear Information System (INIS)

E.N. Lindner

2004-01-01

The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

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)

Comparison of seismic isolation concepts for FBR

International Nuclear Information System (INIS)

Shiojiri, H.; Mazda, T.; Kasai, H.; Kanda, J.N.; Kubo, T.; Madokoro, M.; Shimomura, T.; Nojima, O.

1989-01-01

This paper seeks to verify the reliability and effectiveness of seismic isolation for FBR. Some results of the preliminary study of the program are described. Seismic isolation concepts and corresponding seismic isolation devices were selected. Three kinds of seismically-isolated FBR plant concepts were developed by applying promising seismic isolation concepts to the non-isolated FBR plant, and by developing plant component layout plans and building structural designs. Each plant was subjected to seismic response analysis and reduction in the amount of material of components and buildings were estimated for each seismic isolation concepts. Research and development items were evaluated

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

International Nuclear Information System (INIS)

Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.

1994-01-01

This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)

Seismic structural fragility investigation for the Zion Nuclear Power Plant. Seismic safety margins research program (phase 1)

International Nuclear Information System (INIS)

Wesley, D.A.; Hashimoto, P.S.

1981-10-01

An evaluation of the seismic capacity of the essential structures for the Zion Nuclear Power Plant in Zion, Illinois, was conducted as part of the Seismic Safety Margins Research Program (SSMRP). The structures included the reactor containment building, the turbine/auxiliary building, and the crib house (intake structure). The evaluation was devoted to seismically induced failures rather than those resulting from combined Loss of Coolant Accident (LOCA) or other extreme load combinations. The seismic loads used in the investigation were based on elastic analyses. The loads for the reactor containment and turbine/auxiliary buildings were developed by Lawrence Livermore Laboratory using time history analyses. The loads used for the crib house were the original seismic design loads developed by Sargent and Lundy. No non-linear seismic analyses were conducted. The seismic capacity of the structures accounted for the actual concrete and steel material properties including the aging of the concrete. Median centered properties were used throughout the evaluation including levels of damping considered appropriate for structures close to collapse as compared to the more conservative values used for design. The inelastic effects were accounted for using ductility modified response spectrum techniques based on system ductility ratios expected for structures near collapse. Sources of both inherent randomness and uncertainties resulting from lack of knowledge or approximations in analytical modelling were considered in developing the dispersion of the structural dynamic characteristics. Coefficients of variation were developed assuming lognormal distributions for all variables. The earthquake levels for many of the seismically induced failure modes are so high as to be considered physically incredible. (author)

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)

Seismic Isolation Working Meeting Gap Analysis Report

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-01

The ultimate goal in nuclear facility and nuclear power plant operations is operating safety during normal operations and maintaining core cooling capabilities during off-normal events including external hazards. Understanding the impact external hazards, such as flooding and earthquakes, have on nuclear facilities and NPPs is critical to deciding how to manage these hazards to expectable levels of risk. From a seismic risk perspective the goal is to manage seismic risk. Seismic risk is determined by convolving the seismic hazard with seismic fragilities (capacity of systems, structures, and components (SSCs)). There are large uncertainties associated with evolving nature of the seismic hazard curves. Additionally there are requirements within DOE and potential requirements within NRC to reconsider updated seismic hazard curves every 10 years. Therefore opportunity exists for engineered solutions to manage this seismic uncertainty. One engineered solution is seismic isolation. Current seismic isolation (SI) designs (used in commercial industry) reduce horizontal earthquake loads and protect critical infrastructure from the potentially destructive effects of large earthquakes. The benefit of SI application in the nuclear industry is being recognized and SI systems have been proposed, in the American Society of Civil Engineers (ASCE) 4 standard, to be released in 2014, for Light Water Reactors (LWR) facilities using commercially available technology. However, there is a lack of industry application to the nuclear industry and uncertainty with implementing the procedures outlined in ASCE-4. Opportunity exists to determine barriers associated with implementation of current ASCE-4 standard language.

Framework for estimating response time data to conduct a seismic human reliability analysis - its feasibility

International Nuclear Information System (INIS)

Park, Jinkyun; Kin, Yochan; Jung, Wondea; Jang, Seung Cheol

2014-01-01

This is because the PSA has been used for several decades as the representative tool to evaluate the safety of NPPs. To this end, it is inevitable to evaluate human error probabilities (HEPs) in conducting important tasks being considered in the PSA framework (i.e., HFEs; human failure events), which are able to significantly affect the safety of NPPs. In addition, it should be emphasized that the provision of a realistic human performance data is an important precondition for calculating HEPs under a seismic condition. Unfortunately, it seems that HRA methods being currently used for calculating HEPs under a seismic event do not properly consider the performance variation of human operators. For this reason, in this paper, a framework to estimate response time data that are critical for calculating HEPs is suggested with respect to a seismic intensity. This paper suggested a systematic framework for estimating response time data that would be one of the most critical for calculating HEPs. Although extensive review of existing literatures is indispensable for identifying response times of human operators who have to conduct a series of tasks prescribed in procedures based on a couple of wrong indications, it is highly expected that response time data for seismic HRA can be properly secured through revisiting response time data collected from diverse situations without concerning a seismic event

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

International Nuclear Information System (INIS)

Payne, Suzette Jackson; Coppersmith, Ryan; Coppersmith, Kevin; Rodriguez-Marek, Adrian; Falero, Valentina Montaldo; Youngs, Robert

2016-01-01

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Advanced Test Reactor (ATR), and Naval Reactors Facility (NRF) at the Idaho National Laboratory (INL). The PSHA followed the approaches and procedures for Senior Seismic Hazard Analysis Committee (SSHAC) Level 1 study and included a Participatory Peer Review Panel (PPRP) to provide the confident technical basis and mean-centered estimates of the ground motions. A new risk-informed methodology for evaluating the need for an update of an existing PSHA was developed as part of the Seismic Risk Assessment (SRA) project. To develop and implement the new methodology, the SRA project elected to perform two SSHAC Level 1 PSHAs. The first was for the Fuel Manufacturing Facility (FMF), which is classified as a Seismic Design Category (SDC) 3 nuclear facility. The second was for the ATR Complex, which has facilities classified as SDC-4. The new methodology requires defensible estimates of ground motion levels (mean and full distribution of uncertainty) for its criteria and evaluation process. The INL SSHAC Level 1 PSHA demonstrates the use of the PPRP, evaluation and integration through utilization of a small team with multiple roles and responsibilities (four team members and one specialty contractor), and the feasibility of a short duration schedule (10 months). Additionally, a SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels for the Spent Fuel Handling Recapitalization Project (SFHP) process facility.

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Payne, Suzette Jackson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coppersmith, Ryan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coppersmith, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rodriguez-Marek, Adrian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Falero, Valentina Montaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngs, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Advanced Test Reactor (ATR), and Naval Reactors Facility (NRF) at the Idaho National Laboratory (INL). The PSHA followed the approaches and procedures for Senior Seismic Hazard Analysis Committee (SSHAC) Level 1 study and included a Participatory Peer Review Panel (PPRP) to provide the confident technical basis and mean-centered estimates of the ground motions. A new risk-informed methodology for evaluating the need for an update of an existing PSHA was developed as part of the Seismic Risk Assessment (SRA) project. To develop and implement the new methodology, the SRA project elected to perform two SSHAC Level 1 PSHAs. The first was for the Fuel Manufacturing Facility (FMF), which is classified as a Seismic Design Category (SDC) 3 nuclear facility. The second was for the ATR Complex, which has facilities classified as SDC-4. The new methodology requires defensible estimates of ground motion levels (mean and full distribution of uncertainty) for its criteria and evaluation process. The INL SSHAC Level 1 PSHA demonstrates the use of the PPRP, evaluation and integration through utilization of a small team with multiple roles and responsibilities (four team members and one specialty contractor), and the feasibility of a short duration schedule (10 months). Additionally, a SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels for the Spent Fuel Handling Recapitalization Project (SFHP) process facility.

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.)

Seismic evaluation of existing nuclear facilities. Proceedings of the SMiRT-14 post conference seminar no. 16

International Nuclear Information System (INIS)

2001-03-01

Since 1992 the IAEA has been assisting Member States to develop NPP specific guidelines used in post-construction seismic safety reevaluation. Technical guidelines were developed with the aim to establish a general framework within which a seismic reevaluation of an operating NPP can be carried out. These guidelines will form the basis of an IAEA Safety Report on seismic evaluation of the existing nuclear facilities. Forty papers from 26 countries were presented. Most of the papers reviewed activities in the field of seismic reevaluation and upgrading of existing plants carried out in recent years. This is essential for the IAEA in the development of a unified approach to the seismic reevaluation of existing facilities applicable to WWERs, CANDU, PWRs, etc. A number of papers deal with the CRP benchmark study for the seismic analysis and testing of WWER type NPPs organized by the IAEA (1993-1997). The focal activities of the CRP were benchmarking exercises, applying similar methodology both for Paks NPP and Kozloduy NPP Unit 5 to test the NPP (mainly reactor building) using blast loading generated by a series of artificially generated underground explosions. This work highlighted the reliability of the available numerical tools, the need for further research, and a general judgement on the best compromise between experimental and numerical tools in the seismic reevaluation process. The final results of the CRP were presented in IAEA-TECDOC-1176

Seismic evaluation of existing nuclear facilities. Proceedings of the SMiRT-14 post conference seminar no. 16

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

Since 1992 the IAEA has been assisting Member States to develop NPP specific guidelines used in post-construction seismic safety reevaluation. Technical guidelines were developed with the aim to establish a general framework within which a seismic reevaluation of an operating NPP can be carried out. These guidelines will form the basis of an IAEA Safety Report on seismic evaluation of the existing nuclear facilities. Forty papers from 26 countries were presented. Most of the papers reviewed activities in the field of seismic reevaluation and upgrading of existing plants carried out in recent years. This is essential for the IAEA in the development of a unified approach to the seismic reevaluation of existing facilities applicable to WWERs, CANDU, PWRs, etc. A number of papers deal with the CRP benchmark study for the seismic analysis and testing of WWER type NPPs organized by the IAEA (1993-1997). The focal activities of the CRP were benchmarking exercises, applying similar methodology both for Paks NPP and Kozloduy NPP Unit 5 to test the NPP (mainly reactor building) using blast loading generated by a series of artificially generated underground explosions. This work highlighted the reliability of the available numerical tools, the need for further research, and a general judgement on the best compromise between experimental and numerical tools in the seismic reevaluation process. The final results of the CRP were presented in IAEA-TECDOC-1176.

IEEE C37.98-1978: IEEE standard seismic testing of relays

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard specifies the procedures to be used in the seismic testing of relays used in power system facilities. The standard is concerned with the determination of the seismic fragility level of relays and also gives recommendations for proof testing. The purpose of this standard is to establish procedures for determining the seismic capabilities of protective and auxiliary relays. These procedures employ what has been called fragility testing in ANSI/IEEE Std 344-1975, Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations. In order to define the conditions for fragility testing of relays, parameters in three separate areas must be specified. In general they are: (1) the electrical settings and inputs to the relay, and other information to define its conditions during the test; (2) the change in state, deviation in operating characteristics or tolerances, or other change of performance of the relay which constitutes failure; (3) the seismic vibration environment to be imposed during the test. Since it is not possible to define the conditions for every conceivable application for all relays, those parameters, which in practice encompass the majority of applications, have been specified in this standard. When the application of the relay is other than as specified under any of (1), (2), and (3), or if it is not practical to apply existing results of fragility tests to that new case

Evaluation of methods for seismic analysis of nuclear fuel reprocessing plants, part 1

International Nuclear Information System (INIS)

Tokarz, F.J.; Murray, R.C.; Arthur, D.F.; Feng, W.W.; Wight, L.H.; Zaslawsky, M.

1975-01-01

Currently, no guidelines exist for choosing methods of structural analysis to evaluate the seismic hazard of nuclear fuel reprocessing plants. This study examines available methods and their applicability to fuel reprocessing plant structures. The results of this study should provide a basis for establishing guidelines recommending methods of seismic analysis for evaluating future fuel reprocessing plants. The approach taken is: (1) to identify critical plant structures and place them in four categories (structures at or near grade; deeply embedded structures; fully buried structures; equipment/vessels/attachments/piping), (2) to select a representative structure in each of the first three categories and perform static and dynamic analysis on each, and (3) to evaluate and recommend method(s) of analysis for structures within each category. The Barnwell Nuclear Fuel Plant is selected as representative of future commercial reprocessing plants. The effect of site characteristics on the structural response is also examined. The response spectra method of analysis combined with the finite element model for each category is recommended. For structures founded near or at grade, the lumped mass model could also be used. If a time history response is required, a time-history analysis is necessary. (U.S.)

Input parameters for the statistical seismic hazard assessment in central part of Romania territory using crustal earthquakes

International Nuclear Information System (INIS)

Moldovan, A.I.; Bazacliu, O.; Popescu, E.

2004-01-01

The seismic hazard assessment in dense-populated geographical regions and subsequently the design of the strategic objectives (dams, nuclear power plants, etc.) are based on the knowledge of the seismicity parameters of the seismic sources which can generate ground motion amplitudes above the minimum level considered risky at the specific site and the way the seismic waves propagate between the focus and the site. The purpose of this paper is to provide a set of information required for a probabilistic assessment of the seismic hazard in the central Romanian territory relative to the following seismic sources: Fagaras zone (FC), Campulung zone (CP), and Transilvania zone (TD) all of them in the crust domain. Extremely vulnerable objectives are present in the central part of Romania, including cities of Pitesti and Sibiu and the 'Vidraru' dam. The analysis that we propose implies: (1) geometrical definition of the seismic sources, (2) estimation of the maximum possible magnitude, (3) estimation of the frequency - magnitude relationship and (4) estimation of the attenuation laws. As an example, the obtained input parameters are used to evaluate the seismic hazard distribution due to the crustal earthquakes applying the McGuire's procedure (1976). These preliminary results are in good agreement with the previous research based on deterministic approach (Radulian et al., 2000). (authors)

Strong Motion Instrumentation of Seismically-Strengthened Port Structures in California by CSMIP

Science.gov (United States)

Huang, M.J.; Shakal, A.F.

2009-01-01

The California Strong Motion Instrumentation Program (CSMIP) has instrumented five port structures. Instrumentation of two more port structures is underway and another one is in planning. Two of the port structures have been seismically strengthened. The primary goals of the strong motion instrumentation are to obtain strong earthquake shaking data for verifying seismic analysis procedures and strengthening schemes, and for post-earthquake evaluations of port structures. The wharves instrumented by CSMIP were recommended by the Strong Motion Instrumentation Advisory Committee, a committee of the California Seismic Safety Commission. Extensive instrumentation of a wharf is difficult and would be impossible without the cooperation of the owners and the involvement of the design engineers. The instrumentation plan for a wharf is developed through study of the retrofit plans of the wharf, and the strong-motion sensors are installed at locations where specific instrumentation objectives can be achieved and access is possible. Some sensor locations have to be planned during design; otherwise they are not possible to install after construction. This paper summarizes the two seismically-strengthened wharves and discusses the instrumentation schemes and objectives. ?? 2009 ASCE.

New Technology Changing The Face of Mobile Seismic Networks

Science.gov (United States)

Brisbourne, A.; Denton, P.; Seis-Uk

SEIS-UK, a seismic equipment pool and data management facility run by a consortium of four UK universities (Leicester, Leeds, Cambridge and Royal Holloway, London) completed its second phase in 2001. To compliment the existing broadband equipment pool, which has been deployed to full capacity to date, the consortium undertook a tender evaluation process for low-power, lightweight sensors and recorders, for use on both controlled source and passive seismic experiments. The preferred option, selected by the consortium, was the Guralp CMG-6TD system, with 150 systems ordered. The CMG-6TD system is a new concept in temporary seismic equipment. A 30s- 100Hz force-feedback sensor, integral 24bit digitiser and 3-4Gbyte of solid-state memory are all housed in a single unit. Use of the most recent technologies has kept the power consumption to below 1W and the weight to 3.5Kg per unit. The concept of the disk-swap procedure for obtaining data from the field has been usurped by a fast data download technique using firewire technology. This allows for rapid station servicing, essential when 150 stations are in use, and also ensures the environmental integrity of the system by removing the requirement for a disk access port and envi- ronmentally exposed data disk. The system therefore meets the criteria for controlled source and passive seismic experiments: (1) the single unit concept and low-weight is designed for rapid deployment on short-term projects; (2) the low power consumption reduces the power-supply requirements facilitating deployment; (3) the low self-noise and bandwidth of the sensor make it applicable to passive experiments involving nat- ural sources. Further to this acquisition process, in collaboration with external groups, the SEIS- UK data management procedures have been streamlined with the integration of the Guralp GCF format data into the PASSCAL PDB software. This allows for rapid dissemination of field data and the production of archive-ready datasets

Study on seismic stability of seawall in man-made island. Pt. 3. Development of evaluation method for seismic stability of seawall considering dynamic interaction of armour embankment and caisson

International Nuclear Information System (INIS)

Tochigi, Hitoshi; Kanatani, Mamoru; Kawai, Tadashi

1999-01-01

In the development of siting technology for off-shore nuclear power plants on man-made island, assessing the stability of seawall which ensure the safety of backfill ground against ocean waves and earthquakes is indispensable. Concerning the seismic stability of seawall, it is indicated that the evaluation of the influence of lateral pressure of armour embankment acting on the caisson wall is an important factor. We proposed modeling method of discontinuity such as armour unit by distinct element method (DEM) and combined them with caisson type seawall model idealized by FEM. To show the validity of this DEM-FEM hybrid analysis method, the numerical simulation for the shaking table tests was conducted. For the two kinds of seawall model with armour embankment and without the one, the lateral displacement of the seawall induced by sliding of caisson is successfully reproduced and accuracy of above mentioned numerical procedure is confirmed. (author)

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.

Nonlinear seismic analysis of reinforced concrete framed structures considering joint distortion

International Nuclear Information System (INIS)

Sharma, Akanshu; Reddy, G.R.; Vaze, K.K.; Eligehausen, Rolf; Hofmann, J.

2012-01-01

Seismic behavior of a reinforced concrete framed structure can be assessed with various analytical tools that may broadly be classified as linear elastic procedures and non-linear or inelastic analysis procedures. Since the reinforced concrete structures generally go in the inelastic range due to seismic loading, it can be easily said that the inelastic procedures would predict the performance of the structures in a much better and realistic way than the linear elastic procedures. However, at the same time, the inelastic procedures are computationally much more demanding. Thus, a good balance between accuracy and computational effort is often sought for. To assess the seismic behaviour of reinforced concrete framed structures, various experimental procedures can be used. Pushover tests that consist of loading the structure monotonically till failure can be conducted on large scale structures and give information about the load carrying and deformational capacity of the structure along with sequence of failure modes but only in one direction. Static cyclic tests, where inertia effects are not included give the above mentioned information for to and fro loading direction along with the information on energy consumption. Shake table tests, which are closest to the real life earthquake tests provide almost all the information required to understand the seismic behaviour but the scale of such tests are usually limited by the capacity of the shaking table facility. In this work, practically usable and sufficiently accurate models are reported to realistically model the inelastic response of the structures. A new model to consider the inelastic behaviour of the joints of poorly detailed structures is developed and presented. A practical hysteretic rule based on the extension of Pivot hysteretic model is developed for members and beam-column joints and the same is also reported. The analytical models are validated against the experimental results using pushover analysis

Application of the SQUG-GIP to the seismic upgrade program of the Savannah River reactors

International Nuclear Information System (INIS)

Antaki, G.A.

1991-01-01

In August 1991, the Savannah River Site (SRS) seismic evaluation program using the Generic Implementation Procedure (GIP) celebrated its third anniversary-a respectable age for such a new methodology. During these three years, the GIP, developed for the commercial nuclear industry's Seismic Qualification Utility Group (SQUG), had evolved through Revision 01, Revision 1, Revision 2 and a Revision 2 open-quotes updateclose quotes which is currently in the works. This evolution is not surprising for such an important, and in many ways pioneering, document. The various revisions were anticipated at SRS, and the program adjusted accordingly. The verification of seismic adequacy of equipment at the SRS nuclear reactors has been outlined in previous publications. The purpose of this paper is to relate the more practical and managerial aspects of our relatively mature SQUG-GIP implementation program, which will hopefully prove useful to future users of the GIP. This report is divided into four sections, which follow the normal flow of work under GIP: (1) Program Prerequisites; (2) Definition of Scope; (3) Equipment Evaluations; and (4) Resolution of Outliers

Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

International Nuclear Information System (INIS)

Luk, V.K.; Spencer, B.W.; Shaukat, S.K.; Lam, I.P.; Dameron, R.A.

2003-01-01

Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

Calibration of Seismic Attributes for Reservoir Characterization

Energy Technology Data Exchange (ETDEWEB)

Pennington, Wayne D.; Acevedo, Horacio; Green, Aaron; Len, Shawn; Minavea, Anastasia; Wood, James; Xie, Deyi

2002-01-29

This project has completed the initially scheduled third year of the contract, and is beginning a fourth year, designed to expand upon the tech transfer aspects of the project. From the Stratton data set, demonstrated that an apparent correlation between attributes derived along `phantom' horizons are artifacts of isopach changes; only if the interpreter understands that the interpretation is based on this correlation with bed thickening or thinning, can reliable interpretations of channel horizons and facies be made. From the Boonsville data set , developed techniques to use conventional seismic attributes, including seismic facies generated under various neural network procedures, to subdivide regional facies determined from logs into productive and non-productive subfacies, and developed a method involving cross-correlation of seismic waveforms to provide a reliable map of the various facies present in the area. The Teal South data set provided a surprising set of data, leading us to develop a pressure-dependent velocity relationship and to conclude that nearby reservoirs are undergoing a pressure drop in response to the production of the main reservoir, implying that oil is being lost through their spill points, never to be produced. The Wamsutter data set led to the use of unconventional attributes including lateral incoherence and horizon-dependent impedance variations to indicate regions of former sand bars and current high pressure, respectively, and to evaluation of various upscaling routines.

Vessel core seismic interaction for a fast reactor

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1984-01-01

This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

The seismic monitoring network of Mt. Vesuvius

Directory of Open Access Journals (Sweden)

Massimo Orazi

2013-11-01

Full Text Available Mt. Vesuvius (southern Italy is one of the most hazardous volcanoes in the world. Its activity is currently characterized by moderate seismicity, with hypocenters located beneath the crater zone with depth rarely exceeding 5 km and magnitudes generally less than 3. The current configuration of the seismic monitoring network of Mt. Vesuvius consists of 18 seismic stations and 7 infrasound microphones. During the period 2006-2010 a seismic array with 48 channels was also operative. The station distribution provides appropriate coverage of the area around the volcanic edifice. The current development of the network and its geometry, under conditions of low seismic noise, allows locating seismic events with M<1. Remote instruments continuously transmit data to the main acquisition center in Naples. Data transmission is realized using different technological solutions based on UHF, Wi-Fi radio links, and TCP/IP client-server applications. Data are collected in the monitoring center of the Osservatorio Vesuviano (Italian National Institute of Geophysics and Volcanology, Naples section, which is equipped with systems for displaying and analyzing signals, using both real-time automatic and manual procedures. 24-hour surveillance allows to immediately communicate any significant anomaly to the Civil Protection authorities.

Seismic performance evaluation of high natural frequency mechanical structure from the viewpoint of energy balance

International Nuclear Information System (INIS)

Minagawa, Keisuke; Fujita, Satoshi; Endo, Rokuro; Amemiya, Mitsuhiko

2009-01-01

In this study, vibration characteristics of mechanical structure having high natural frequency are investigated from the viewpoint of energy balance. Mechanical structures having high natural frequency in a nuclear power plant are generally designed statically and elastically. However it has been reported that fracture of ordinary piping is produced not by momentary large load but by cumulative fatigue damage. Therefore it is very important to grasp seismic performance dynamically by considering cyclic load. This paper deals with an investigation regarding seismic performance evaluation of high natural frequency mechanical structure. The energy balance equation that is one of valid methods for structural calculation is applied through the investigation. The main feature of the energy balance equation is that it explains accumulated information of motion. Therefore the energy balance equation is adequate for the investigation of the influence of cumulative load such as seismic response. In this paper, vibration experiment and simulation using sinusoidal waves and artificial seismic waves were examined in order to investigate relationship between natural frequency of structure and energy. As a result, we found that input energy decreases with an increase in the natural frequency. (author)

Use of response envelopes for seismic margin assessment of reinforced concrete walls and slabs

Energy Technology Data Exchange (ETDEWEB)

Ile, Nicolas; Frau, Alberto, E-mail: alberto.frau@cea.fr

2017-04-01

Highlights: • Proposal of a method for application of the elliptical envelope to RC shell elements. • Proposal of new algorithms for the seismic margin evaluation for RC shell elements. • Verification of a RC wall 3D structure, using the proposed assessment approach. - Abstract: Seismic safety evaluations of existing nuclear facilities are usually based on the assumption of structural linearity. For the design basis earthquake (DBE), it is reasonable to apply a conventional evaluation of the seismic safety of building structures and carry out a linear elastic analysis to assess the load effects on structural elements. Estimating the seismic capacity of a structural element requires an estimation of the critical combination of responses acting in this structural element and compare this combination with the capacity of the element. By exploiting the response-spectrum-based procedure for predicting the response envelopes in linear structures formulated by Menun and Der Kiureghian (2000a), algorithms are developed for the seismic margin assessment of reinforced concrete shell finite elements. These algorithms facilitate the comparison of the response-spectrum-based envelopes to prescribed capacity surfaces for the purpose of assessing the safety margin of this kind of structures. The practical application of elliptical response envelopes in case of shell finite elements is based on the use of layer models such as those developed by Marti (1990), which transfer the generalized stress field to three layers under the assumption that the two outer layers carry membrane forces and the internal layer carries only the out-of-plane shears. The utility of the assessment approach is discussed with reference to a case study of a 3D structure made of reinforced concrete walls.

Use of response envelopes for seismic margin assessment of reinforced concrete walls and slabs

International Nuclear Information System (INIS)

Ile, Nicolas; Frau, Alberto

2017-01-01

Highlights: • Proposal of a method for application of the elliptical envelope to RC shell elements. • Proposal of new algorithms for the seismic margin evaluation for RC shell elements. • Verification of a RC wall 3D structure, using the proposed assessment approach. - Abstract: Seismic safety evaluations of existing nuclear facilities are usually based on the assumption of structural linearity. For the design basis earthquake (DBE), it is reasonable to apply a conventional evaluation of the seismic safety of building structures and carry out a linear elastic analysis to assess the load effects on structural elements. Estimating the seismic capacity of a structural element requires an estimation of the critical combination of responses acting in this structural element and compare this combination with the capacity of the element. By exploiting the response-spectrum-based procedure for predicting the response envelopes in linear structures formulated by Menun and Der Kiureghian (2000a), algorithms are developed for the seismic margin assessment of reinforced concrete shell finite elements. These algorithms facilitate the comparison of the response-spectrum-based envelopes to prescribed capacity surfaces for the purpose of assessing the safety margin of this kind of structures. The practical application of elliptical response envelopes in case of shell finite elements is based on the use of layer models such as those developed by Marti (1990), which transfer the generalized stress field to three layers under the assumption that the two outer layers carry membrane forces and the internal layer carries only the out-of-plane shears. The utility of the assessment approach is discussed with reference to a case study of a 3D structure made of reinforced concrete walls.

Structural analysis and optimization procedure of the TFTR device substructure

International Nuclear Information System (INIS)

Driesen, G.

1975-10-01

A structural evaluation of the TFTR device substructure is performed in order to verify the feasibility of the proposed design concept as well as to establish a design optimization procedure for minimizing the material and fabrication cost of the substructure members. A preliminary evaluation of the seismic capability is also presented. The design concept on which the analysis is based is consistent with that described in the Conceptual Design Status Briefing report dated June 18, 1975

A Site Characterization Protocol for Evaluating the Potential for Triggered or Induced Seismicity Resulting from Wastewater Injection and Hydraulic Fracturing

Science.gov (United States)

Walters, R. J.; Zoback, M. D.; Gupta, A.; Baker, J.; Beroza, G. C.

2014-12-01

Regulatory and governmental agencies, individual companies and industry groups and others have recently proposed, or are developing, guidelines aimed at reducing the risk associated with earthquakes triggered by waste water injection or hydraulic fracturing. While there are a number of elements common to the guidelines proposed, not surprisingly, there are also some significant differences among them and, in a number of cases, important considerations that are not addressed. The goal of this work is to develop a comprehensive protocol for site characterization based on a rigorous scientific understanding of the responsible processes. Topics addressed will include the geologic setting (emphasizing faults that might be affected), historical seismicity, hydraulic characterization of injection and adjacent intervals, geomechanical characterization to identify potentially active faults, plans for seismic monitoring and reporting, plans for monitoring and reporting injection (pressure, volumes, and rates), other factors contributing to risk (potentially affected population centers, structures, and facilities), and implementing a modified Probabilistic Seismic Hazard Analysis (PSHA). The guidelines will be risk based and adaptable, rather than prescriptive, for a proposed activity and region of interest. They will be goal oriented and will rely, to the degree possible, on established best practice procedures, referring to existing procedures and recommendations. By developing a risk-based site characterization protocol, we hope to contribute to the development of rational and effective measures for reducing the risk posed by activities that potentially trigger earthquakes.

Existing reflection seismic data re-processing

International Nuclear Information System (INIS)

Higashinaka, Motonori; Sano, Yukiko; Kozawa, Takeshi

2005-08-01

This document is to report the results of existing seismic data re-processing around Horonobe town, Hokkaido, Japan, which is a part of the Horonobe Underground Research Project. The main purpose of this re-processing is to recognize the subsurface structure of Omagari Fault and fold system around Omagari Fault. The seismic lines for re-processing are TYHR-A3 line and SHRB-2 line, which JAPEX surveyed in 1975. Applying weathering static correction using refraction analysis and noise suppression procedure, we have much enhanced seismic profile. Following information was obtained from seismic re-processing results. TYHR-A3 line: There are strong reflections, dipping to the west. These reflections are corresponding western limb of anticline to the west side of Omagari Fault. SHRB-2 line: There are strong reflections, dipping to the west, at CDP 60-140, while there are reflections, dipping to the east, to the east side of CDP 140. These reflections correspond to the western limb and the eastern limb of the anticline, which is parallel to Omagari FAULT. This seismic re-processing provides some useful information to know the geological structure around Omagari Fault. (author)

Seismic hazard for the Savannah River Site: A comparative evaluation of the EPRI and LLNL assessments

International Nuclear Information System (INIS)

Wingo, H.E.

1992-01-01

This report was conducted to: (1) develop an understanding of causes for the vast differences between the two comprehensive studies, and (2) using a methodology consistent with the reconciled methods employed in the two studies, develop a single seismic hazard for the Savannah River Site suitable for use in seismic probabilistic risk assessments with emphasis on the K Reactor. Results are presented for a rock site which is a typical because detailed evaluations of soil characteristics at the K Reactor are still in progress that account for the effects of a soil stablizing grouting program. However when the soils analysis is completed, the effects of soils can be included with this analysis with the addition of a single factor that will decrease slightly the seismic hazard for a rock site

Automated control procedures and first results from the temporary seismic monitoring of the 2012 Emilia sequence

Directory of Open Access Journals (Sweden)

Simone Marzorati

2012-10-01

Full Text Available After moderate to strong earthquakes in Italy or in the surrounding areas, the Istituto Nazionale di Geofisica e Vulcanologia (INGV; National Institute for Geophysics and Volcanology activates a temporary seismic network infrastructure. This is devoted to integration with the Italian National Seismic Network (RSN [Delladio 2011] in the epicentral area, thus improving the localization of the aftershocks distribution after a mainshock. This infrastructure is composed of a stand-alone, locally recording part (Re.Mo. [Moretti et al. 2010] and a real-time telemetered part (Re.Mo.Tel. [Abruzzese et al. 2011a, 2011b] that can stream data to the acquisition centers in Rome and Grottaminarda. After the May 20, 2012, Ml 5.9 earthquake in the Emilia region (northern Italy, the temporary network was deployed in the epicentral area; in particular, 10 telemetered and 12 stand-alone stations were installed [Moretti et al. 2012, this volume]. Using the dedicated connection between the acquisition center in Rome and the Ancona acquisition sub-center [Cattaneo et al. 2011], the signals of the real-time telemetered stations were acquired also in this sub-center. These were used for preliminary quality control, by adopting the standard procedures in use here (see next paragraph, and Monachesi et al. [2011]. The main purpose of the present study is a first report on this quality check, which should be taken into account for the correct use of these data. […

A quick seismic assessment method for jacket type offshore structures by combining push-over and nonlinear time history analyses

Energy Technology Data Exchange (ETDEWEB)

Karimiyan, S.; Hosseini, M. [International Inst. of Earthquake Engineering and Seismology, Tehran (Iran, Islamic Republic of); Karimiyan, M. [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Earthquake Eng. Dept., School of Engineering

2010-07-01

Several offshore structures are located in seismic regions. In order to upgrade their seismic behaviour, their seismic vulnerability must be evaluated. It is thought that the most reliable type of analysis for seismic evaluation is nonlinear time history analysis (NLTHA), however, it is known to be a very time consuming method. This paper presented a quick procedure by combining the push over analysis (POA) and the NLTHA. The paper discussed both methods in detail. In order to identify the more critical members of the structure, based on the range of their plastic deformations, some POA were first performed. The NLTHA was then performed, focusing on the critical members, to obtain their vulnerability with higher reliability. An offshore structure of jacket type, installed in the Lavan oil field in the Persian Gulf in 1970, was also considered in order to demonstrate the efficiency of the proposed method. It was concluded from the numerical results that combining POA and NLTHA was a quick and reliable seismic evaluation method. The results demonstrated that although the vulnerability of the jacket structure was not very high, the level of damage was not the same for different members, and was dependent on their location in the structure and also its geometric orientation and load bearing situation. 6 refs., 1 tab., 8 figs.

Approaches that use seismic hazard results to address topics of nuclear power plant seismic safety, with application to the Charleston earthquake issue

International Nuclear Information System (INIS)

Sewell, R.T.; McGuire, R.K.; Toro, G.R.; Stepp, J.C.; Cornell, C.A.

1990-01-01

Plant seismic safety indicators include seismic hazard at the SSE (safe shut-down earthquake) acceleration, seismic margin, reliability against core damage, and reliability against offsite consequences. This work examines the key role of hazard analysis in evaluating these indicators and in making rational decisions regarding plant safety. The paper outlines approaches that use seismic hazard results as a basis for plant seismic safety evaluation and applies one of these approaches to the Charleston earthquake issue. This approach compares seismic hazard results that account for the Charleston tectonic interpretation, using the EPRI-Seismicity Owners Group (SOG) methodology, with hazard results that are consistent with historical tectonic interpretations accepted in regulation. Based on hazard results for a set of 21 eastern U.S. nuclear power plant sites, the comparison shows that no systematic 'plant-to-plant' increase in hazard accompanies the Charleston hypothesis; differences in mean hazards for the two interpretations are generally insignificant relative to current uncertainties in seismic hazard. (orig.)

GUI program to compute probabilistic seismic hazard analysis

International Nuclear Information System (INIS)

Shin, Jin Soo; Chi, H. C.; Cho, J. C.; Park, J. H.; Kim, K. G.; Im, I. S.

2005-12-01

The first stage of development of program to compute probabilistic seismic hazard is completed based on Graphic User Interface (GUI). The main program consists of three part - the data input processes, probabilistic seismic hazard analysis and result output processes. The first part has developed and others are developing now in this term. The probabilistic seismic hazard analysis needs various input data which represent attenuation formulae, seismic zoning map, and earthquake event catalog. The input procedure of previous programs based on text interface take a much time to prepare the data. The data cannot be checked directly on screen to prevent input erroneously in existing methods. The new program simplifies the input process and enable to check the data graphically in order to minimize the artificial error within the limits of the possibility

Preliminary seismic analysis of an innovative near term reactor: Methodology and application

International Nuclear Information System (INIS)

Lo Frano, R.; Pugliese, G.; Forasassi, G.

2010-01-01

Nuclear power plant (NPP) design is strictly dependent on seismic hazard and safety aspects concerned with the external events of the site. Earthquake resistant structures design requires realistic and accurate physical and theoretical models to describe the response of the nuclear power plants (NPPs) that depend on both the ground motion characteristics and the dynamic properties of the structures themselves. In order to improve the design of new NPPs and, at the same time, to retrofit existing ones the dynamic behaviour of structures subjected to critical seismic excitations that may occur during their expected service life must be evaluated. The aim of this work is to select new effective methods to assess NPPs vulnerability by properly capturing the effects of a safe shutdown earthquake (SSE) event on nuclear structures, like the near term deployment IRIS reactor, and to evaluate the seismic resistance capability of as-built structures systems and components. To attain the purpose a validated deterministic methodology based on an accurate finite element modelling coupled to substructure and time history approaches was employed for studying the overall dynamic behaviour of the NPP relevant components. Moreover the set up three-dimensional model was also validated to evaluate the performance and reliability of the adopted FEM code (mesh refinements and type element influence). This detailed numerical assessment, involving the most widely used finite element numerical codes (MSC.Marc and Ansys, allowed to solve, perform and simulate as accurately as possible the dynamic behaviour of structures which may withstand a lot of more or less complicate structural problems. To evaluate the accuracy and the reliability as well as to determine the related error of the set-up procedure, the obtained seismic analyses results in term of accelerations, propagated from the ground to the auxiliary building systems and components, and displacements were compared highlighting a

Seismic gaps and plate tectonics: seismic potential for major boundaries

Energy Technology Data Exchange (ETDEWEB)

McCann, W R; Nishenko, S P; Sykes, L R; Krause, J

1979-01-01

The theory of plate tectonics provides a basic framework for evaluating the potential for future great earthquakes to occur along major plate boundaries. Along most of the transform and convergent plate boundaries considered in this paper, the majority of seismic slip occurs during large earthquakes, i.e., those of magnitude 7 or greater. The concepts that rupture zones, as delineated by aftershocks, tend to abut rather than overlap, and large events occur in regions with histories of both long-and short-term seismic quiescence are used in this paper to delineate major seismic gaps. The term seismic gap is taken to refer to any region along an active plate boundary that has not experienced a large thrust or strike-slip earthquake for more than 30 years. A region of high seismic potential is a seismic gap that, for historic or tectonic reasons, is considered likely to produce a large shock during the next few decades. The seismic gap technique provides estimates of the location, size of future events and origin time to within a few tens of years at best. The accompanying map summarizes six categories of seismic potential for major plate boundaries in and around the margins of the Pacific Ocean and the Caribbean, South Sandwich and Sunda (Indonesia) regions for the next few decades. These six categories are meant to be interpreted as forecasts of the location and size of future large shocks and should not be considered to be predictions in which a precise estimate of the time of occurrence is specified. The categories of potential assigned here provide a rationale for assigning priorities for instrumentation, for future studies aimed at predicting large earthquakes and for making estimates of tsunami potential.

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.)

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.)

Strategy for seismic upgrading of chemical plant taking productivity as criterion of judgment

International Nuclear Information System (INIS)

Oshima, M.; Kase, T.; Yashiro, H.; Fukushima, S.

2005-01-01

Seismic upgrading and modification of existing chemical plant facilities have been performed by means of a procedure of the Seismic Design Code and Guidelines of High-pressure Gas Facilities in Japan. Main purpose of this seismic design code is to ensure public safety at seismic events. From the viewpoints of seismic risk of corporate management, CSR (Corporate Social Responsibility) and productivity of the plants are also important for seismic assessment. In this paper, authors proposed strategy for seismic assessment to select appropriate pre-earthquake upgrading and modification considering productivity of plants based on fault tree analysis. This assessment will enable to select weak damage modes and to allocate countermeasure cost optimally to the selected damage modes. (authors)

Seismic analysis for the ALMR

International Nuclear Information System (INIS)

Tajirian, F.F.

1992-01-01

The Advanced Liquid Metal Reactor (ALMR) design uses seismic isolation as a cost effective approach for simplifying seismic design of the reactor module, and for enhancing margins to handle beyond design basis earthquakes (BDBE). A comprehensive seismic analysis plan has been developed to confirm the adequacy of the design and to support regulatory licensing activities. In this plan state-of-the-art computer programs are used to evaluate the system response of the ALMR. Several factors that affect seismic response will be investigated. These include variability in the input earthquake mechanism, soil-structure interaction effects, and nonlinear response of the isolators. This paper reviews the type of analyses that are planned, and discuses the approach that will be used for validating the specific features of computer programs that are required in the analysis of isolated structures. To date, different linear and nonlinear seismic analyses have been completed. The results of recently completed linear analyses have been summarized elsewhere. The findings of three-dimensional seismic nonlinear analyses are presented in this paper. These analyses were performed to evaluate the effect of changes of isolator horizontal stiffness with horizontal displacement on overall response, to develop an approach for representing BDBE events with return periods exceeding 10,000 years, and to assess margins in the design for BDBEs. From the results of these analyses and bearing test data, it can be concluded that a properly designed and constructed seismic isolation system can accommodate displacements several times the design safe shutdown earthquake (SSE) for the ALMR. (author)

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

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

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

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

Multicomponent seismic applications in coalbed methane development

Energy Technology Data Exchange (ETDEWEB)

Lawton, D.; Trend, S. [Calgary Univ., AB (Canada). Dept. of Geology and Geophysics

2004-07-01

Seismic applications for coalbed methane (CBM) development are used to address the following challenges: lateral continuity of coal zones; vertical continuity of coal seams; permeability of cleats and fractures; coal quality and gas content; wet versus dry coal zones; and, monitoring storage of greenhouse gases. This paper presented a brief description of existing seismic programs, including 2-D and 3-D surface seismic surveys; multicomponent seismic surveys; vertical seismic profiles; cross-well seismic surveys; and, time-lapse seismic surveys. A comparative evaluation of their use in the Horseshoe Canyon Formation and the Ardley Formation was presented. The study showed that variations in reservoir properties resulting from gas production and dewatering can be effectively imaged using seismic surveys. Seismic surveys are useful in reservoir management, monitoring sweep efficiency during enhanced natural gas from coal (NGC) production, monitoring disposal of produced water and verifying storage of carbon dioxide for carbon credits. tabs., figs.

On anti-earthquake design procedure of equipment and pipings in near future

International Nuclear Information System (INIS)

Shibata, H.

1981-01-01

The requirement of anti-earthquake design of nuclear power plants is getting severe year by year. The author will try to discuss how to control its severity and how to find a proper design procedure for licensing of new plants under such severe requirements. On the other hand we suffered from the enormous volumes of documents. To decrease such volumes, the format of documents should be standardized as well as the design procedure standardization. Starting from this point, we need the research and development on the following subjects: i) Standardization of design procedure. ii) Standardization of document. iii) Establishment of standard review procedure using computer. iv) Standardization of earthquake-resistant designed equipment. v) Standardization of anti-earthquake design procedure of piping systems. vi) Introducing margin evaluation procedure to design procedure. vii) Introducing proving test procedure of active component to design procedure. viii) Establishment of evaluation of human reliability in design, fabrication, inspection procedures. ix) Establishment of the proper relation of seismic trigger level and post-earthquake design procedures. (orig./HP)

Approach of seismic upgrading in Kashiwazaki-Kariwa Nuclear Power Plant

International Nuclear Information System (INIS)

Sato, Hitoshi

2009-01-01

Because guide for reviewing seismic design of nuclear power reactor facilities was reworked in 2006, we formulated new Design Base Seismic Motion Ss, and we are doing evaluation of seismic safety (back-check). In Japan, depending on aseismatic importance, equipments are classified into S-class, B-class and C-class. For S-class equipments, we evaluate it on the basis of new Ss, and do seismic upgrading. For B-class and C-class equipments, we do seismic upgrading voluntarily on the basis of the experiences of the Niigataken Chuetsu-Oki (NCO) Earthquake. (author)

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

A guidebook for the operation and maintenance of HANARO seismic monitoring analysis system

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Yoon, Doo Byung; Kim, Hyung Kyoo

2003-09-01

Systems and structures related to HANARO safety are classified as seismic category I. Since 1995, the seismic monitoring system has been utilized for monitoring an earthquake at the HANARO site. The existing seismic monitoring system consists of field sensors and monitoring panel. The analog-type monitoring system with magnetic tape recorder is out-of-date model. In addition, the disadvantage of the existing system is that it does not include signal-analyzing equipment. Therefore, we have improved the analog seismic monitoring system into a new digital Seismic Monitoring Analysis System(SMAS) that can offer precise and detail information of the earthquake signals. This newly developed SMAS is operating at the HANARO instrument room to acquire and analyze the signal of an earthquake. This document is a guidebook for the operation and maintenance of the SMAS. The first chapter gives an outline of the SMAS. The second chapter describes functional capability and specification of the hardware. Chapters 3 and 4 describe starting procedure of the SMAS and how to operate the seismic monitoring program, respectively. Chapter 5 illustrates the seismic analysis algorithm used in the SMAS. The way of operating the seismic analysis program is described in chapter 6. Chapter 7 illustrates the calibration procedure for data acquisition module. Chapter 8 describes the symptoms of common malfunctions and its countermeasure suited to the occasions.

Evaluation of Fourier integral. Spectral analysis of seismic events

International Nuclear Information System (INIS)

Chitaru, Cristian; Enescu, Dumitru

2003-01-01

Spectral analysis of seismic events represents a method for great earthquake prediction. The seismic signal is not a sinusoidal signal; for this, it is necessary to find a method for best approximation of real signal with a sinusoidal signal. The 'Quanterra' broadband station allows the data access in numerical and/or graphical forms. With the numerical form we can easily make a computer program (MSOFFICE-EXCEL) for spectral analysis. (authors)

Impact of ground motion characterization on conservatism and variability in seismic risk estimates

International Nuclear Information System (INIS)

Sewell, R.T.; Toro, G.R.; McGuire, R.K.

1996-07-01

This study evaluates the impact, on estimates of seismic risk and its uncertainty, of alternative methods in treatment and characterization of earthquake ground motions. The objective of this study is to delineate specific procedures and characterizations that may lead to less biased and more precise seismic risk results. This report focuses on sources of conservatism and variability in risk that may be introduced through the analytical processes and ground-motion descriptions which are commonly implemented at the interface of seismic hazard and fragility assessments. In particular, implication of the common practice of using a single, composite spectral shape to characterize motions of different magnitudes is investigated. Also, the impact of parameterization of ground motion on fragility and hazard assessments is shown. Examination of these results demonstrates the following. (1) There exists significant conservatism in the review spectra (usually, spectra characteristic of western U.S. earthquakes) that have been used in conducting past seismic risk assessments and seismic margin assessments for eastern U.S. nuclear power plants. (2) There is a strong dependence of seismic fragility on earthquake magnitude when PGA is used as the ground-motion characterization. When, however, magnitude-dependent spectra are anchored to a common measure of elastic spectral acceleration averaged over the appropriate frequency range, seismic fragility shows no important nor consistent dependence on either magnitude or strong-motion duration. Use of inelastic spectral acceleration (at the proper frequency) as the ground spectrum anchor demonstrates a very similar result. This study concludes that a single, composite-magnitude spectrum can generally be used to characterize ground motion for fragility assessment without introducing significant bias or uncertainty in seismic risk estimates

Impact of ground motion characterization on conservatism and variability in seismic risk estimates

Energy Technology Data Exchange (ETDEWEB)

Sewell, R.T.; Toro, G.R.; McGuire, R.K.

1996-07-01

This study evaluates the impact, on estimates of seismic risk and its uncertainty, of alternative methods in treatment and characterization of earthquake ground motions. The objective of this study is to delineate specific procedures and characterizations that may lead to less biased and more precise seismic risk results. This report focuses on sources of conservatism and variability in risk that may be introduced through the analytical processes and ground-motion descriptions which are commonly implemented at the interface of seismic hazard and fragility assessments. In particular, implication of the common practice of using a single, composite spectral shape to characterize motions of different magnitudes is investigated. Also, the impact of parameterization of ground motion on fragility and hazard assessments is shown. Examination of these results demonstrates the following. (1) There exists significant conservatism in the review spectra (usually, spectra characteristic of western U.S. earthquakes) that have been used in conducting past seismic risk assessments and seismic margin assessments for eastern U.S. nuclear power plants. (2) There is a strong dependence of seismic fragility on earthquake magnitude when PGA is used as the ground-motion characterization. When, however, magnitude-dependent spectra are anchored to a common measure of elastic spectral acceleration averaged over the appropriate frequency range, seismic fragility shows no important nor consistent dependence on either magnitude or strong-motion duration. Use of inelastic spectral acceleration (at the proper frequency) as the ground spectrum anchor demonstrates a very similar result. This study concludes that a single, composite-magnitude spectrum can generally be used to characterize ground motion for fragility assessment without introducing significant bias or uncertainty in seismic risk estimates.

Improvement of seismic observation systems in JOYO

International Nuclear Information System (INIS)

Sumino, Kozo; Suto, Masayoshi; Tanaka, Akihiro

2013-01-01

In the experimental fast reactor 'Joyo' in order to perform the seismic observation in and around the building block and ground, SMAC type seismographs had continuously been used for about 38 years. However, this equipment aged, and the 2011 off the Pacific Coast of Tohoku Earthquake on Mach 11, 2011 increased the importance of seismic data of the reactor facilities from the viewpoint of earthquake-proof safety. For these reasons, Joyo updated the system to the seismic observation system reflecting the latest technology/information, while keeping consistency with the observation data of the former seismographs (SMAC type seismograph). This updating improved various problems on the former observation seismographs. In addition, the installation of now observation points in the locations that are important in seismic safety evaluation expanded the data, and further improved the reliability of the seismic observation and evaluation on 'Joyo'. (A.O.)

Earthquake Ground Motion Measures for Seismic Response Evaluation of Structures

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-15

This study used the assessment results of failure criteria - base shear, story drift, top acceleration and top displacement - for a PSC containment building subjected to 30 sets of near-fault ground motions to evaluate the earthquake ground motion intensity measures. Seven intensity measures, peak ground acceleration(PGA), peak ground velocity(PGV), spectral acceleration(Sa), velocity(Sv), spectrum intensity for acceleration(SIa), velocity(SIv) and displacement(SId), were used to represent alternative ground motion. The regression analyses of the failure criteria for a PSC containment building were carried out to evaluate a proper intensity measure by using two regression models and seven ground motion parameters. The regression analysis results demonstrate the correlation coefficients of the failure criteria in terms of the candidate IM. From the results, spectral acceleration(Sa) is estimated as the best parameter for a evaluation of the structural safety for a seismic PSA.

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

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.

Performance Based Plastic Design of Concentrically Braced Frame attuned with Indian Standard code and its Seismic Performance Evaluation

Directory of Open Access Journals (Sweden)

Sejal Purvang Dalal

2015-12-01

Full Text Available In the Performance Based Plastic design method, the failure is predetermined; making it famous throughout the world. But due to lack of proper guidelines and simple stepwise methodology, it is not quite popular in India. In this paper, stepwise design procedure of Performance Based Plastic Design of Concentrically Braced frame attuned with the Indian Standard code has been presented. The comparative seismic performance evaluation of a six storey concentrically braced frame designed using the displacement based Performance Based Plastic Design (PBPD method and currently used force based Limit State Design (LSD method has also been carried out by nonlinear static pushover analysis and time history analysis under three different ground motions. Results show that Performance Based Plastic Design method is superior to the current design in terms of displacement and acceleration response. Also total collapse of the frame is prevented in the PBPD frame.

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.

Preliminary Seismic Response and Fragility Analysis for DACS Cabinet

Energy Technology Data Exchange (ETDEWEB)

Oh, Jinho; Kwag, Shinyoung; Lee, Jongmin; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

A DACS cabinet is installed in the main control room. The objective of this paper is to perform seismic analyses and evaluate the preliminary structural integrity and seismic capacity of the DACS cabinet. For this purpose, a 3-D finite element model of the DACS cabinet was developed and its modal analyses are carried out to analyze the dynamic characteristics. The response spectrum analyses and the related safety evaluation are then performed for the DACS cabinet subject to seismic loads. Finally, the seismic margin and seismic fragility of the DACS cabinet are investigated. A seismic analysis and preliminary structural integrity of the DACS cabinet under self weight and SSE load have been evaluated. For this purpose, 3-D finite element models of the DACS cabinet were developed. A modal analysis, response spectrum analysis, and seismic fragility analysis were then performed. From the structural analysis results, the DACS cabinet is below the structural design limit of under SSE 0.3g, and can structurally withstand until less than SSE 3g based on an evaluation of the maximum effective stresses. The HCLPF capacity for the DGRS of the SSE 0.3g is 0.55g. A modal analysis, response spectrum analysis, and seismic fragility analysis were then performed. From the structural analysis results, the DACS cabinet is below the structural design limit of under SSE 0.3g, and can structurally withstand until less than SSE 3g based on an evaluation of the maximum effective stresses. The HCLPF capacity for the DGRS of the SSE 0.3g is 0.55g. Therefore, it is concluded that the DACS cabinet was safely designed in that no damage to the preliminary structural integrity and sufficient seismic margin is expected.

Preliminary Seismic Response and Fragility Analysis for DACS Cabinet

International Nuclear Information System (INIS)

Oh, Jinho; Kwag, Shinyoung; Lee, Jongmin; Kim, Youngki

2013-01-01

A DACS cabinet is installed in the main control room. The objective of this paper is to perform seismic analyses and evaluate the preliminary structural integrity and seismic capacity of the DACS cabinet. For this purpose, a 3-D finite element model of the DACS cabinet was developed and its modal analyses are carried out to analyze the dynamic characteristics. The response spectrum analyses and the related safety evaluation are then performed for the DACS cabinet subject to seismic loads. Finally, the seismic margin and seismic fragility of the DACS cabinet are investigated. A seismic analysis and preliminary structural integrity of the DACS cabinet under self weight and SSE load have been evaluated. For this purpose, 3-D finite element models of the DACS cabinet were developed. A modal analysis, response spectrum analysis, and seismic fragility analysis were then performed. From the structural analysis results, the DACS cabinet is below the structural design limit of under SSE 0.3g, and can structurally withstand until less than SSE 3g based on an evaluation of the maximum effective stresses. The HCLPF capacity for the DGRS of the SSE 0.3g is 0.55g. A modal analysis, response spectrum analysis, and seismic fragility analysis were then performed. From the structural analysis results, the DACS cabinet is below the structural design limit of under SSE 0.3g, and can structurally withstand until less than SSE 3g based on an evaluation of the maximum effective stresses. The HCLPF capacity for the DGRS of the SSE 0.3g is 0.55g. Therefore, it is concluded that the DACS cabinet was safely designed in that no damage to the preliminary structural integrity and sufficient seismic margin is expected

Seismic analysis of nuclear power plants

International Nuclear Information System (INIS)

Halbritter, A.L.

1984-01-01

Nuclear Power Plants require exceptional safety guarantees which are reflected in a rigorous control of the employed materials, advanced construction technology, sophisticated methods of analysis and consideration of non conventional load cases such as the earthquake loading. In this paper, the current procedures used in the seismic analysis of Nuclear Power Plants are presented. The seismic analysis of the structures has two objectives: the determination of forces in the structure in order to design it against earthquakes and the generation of floor response spectra to be used in the design of mechanical and electrical components and piping systems. (Author) [pt

Damage assessment for seismic response of recycled concrete filled steel tube columns

Science.gov (United States)

Huang, Yijie; Xiao, Jianzhuang; Shen, Luming

2016-09-01

A model for evaluating structural damage of recycled aggregate concrete filled steel tube (RCFST) columns under seismic effects is proposed in this paper. The proposed model takes the lateral deformation and the effect of repeated cyclic loading into account. Available test results were collected and utilized to calibrate the parameters of the proposed model. A seismic test for six RCFST columns was also performed to validate the proposed damage assessment model. The main test parameters were the recycled coarse aggregate (RCA) replacement percentage and the bond-slip property. The test results indicated that the seismic performance of the RCFST member depends on the RCA contents and their damage index increases as the RCA replacement percentage increases. It is also indicated that the damage degree of RCFST changes with the variation of the RCA replacement percentage. Finally, comparisons between the RCA contents, lateral deformation ratio and damage degree were implemented. It is suggested that an improvement procedure should be implemented in order to compensate for the performance difference between the RCFST and normal concrete filled steel tubes (CFST).

Evaluation of Seismic Hazards at California Department of Transportation (CALTRANS)Structures

Science.gov (United States)

Merriam, M. K.

2005-12-01

The California Department of Transportation (CALTRANS) has responsibility for design, construction, and maintenance of approximately 12,000 state bridges. CALTRANS also provides oversight for similar activities for 12,200 bridges owned by local agencies throughout the state. California is subjected to a M6 or greater seismic event every few years. Recent earthquakes include the 1971 Mw6.6 San Fernando earthquake which struck north of Los Angeles and prompted engineers to begin retrofitting existing bridges and re-examine the way bridges are detailed to improve their response to earthquakes, the 1989 Mw6.9 Loma Prieta earthquake which destroyed the Cypress Freeway and damaged the San Francisco-Oakland Bay Bridge, and the 1994 Mw6.7 Northridge earthquake in the Los Angeles area which heavily damaged four major freeways. Since CALTRANS' seismic performance goal is to ensure life-safety needs are met for the traveling public during an earthquake, estimating earthquake magnitude, peak bedrock acceleration, and determining if special seismic considerationsare needed at specific bridge sites are critical. CALTRANS is currently developing a fourth generation seismic hazard map to be used for estimating these parameters. A deterministic approach has been used to develop this map. Late-Quaternary-age faults are defined as the expected seismic sources. Caltrans requires site-specific studies to determine potential for liquefaction, seismically induced landslides, and surface fault rupture. If potential for one of these seismic hazards exists, the hazard is mitigated by avoidance, removal, or accommodated through design. The action taken, while complying with the Department's "no collapse" requirement, depends upon many factors, including cost.

Romanian seismic network

International Nuclear Information System (INIS)

Ionescu, Constantin; Rizescu, Mihaela; Popa, Mihaela; Grigore, Adrian

2000-01-01

The research in the field of seismology in Romania is mainly carried out by the National Institute for Earth Physics (NIEP). The NIEP activities are mainly concerned with the fundamental research financed by research contracts from public sources and the maintenance and operation of the Romanian seismic network. A three stage seismic network is now operating under NIEP, designed mainly to monitor the Vrancea seismic region in a magnitude range from microearthquakes to strong events: - network of 18 short-period seismometers (S13); - Teledyne Geotech Instruments (Texas); - network of 7 stations with local digital recording (PCM-5000) on magnetic tape, made up of, S13 geophone (T=2 s) on vertical component and SH1 geophone (T=5 s) on horizontal components; - network of 28 SMA-1 accelerometers and 30 digital accelerometers (Kinemetrics - K2) installed in the free field conditions in the framework of the joint German-Romanian cooperation program (CRC); the K2 instruments cover a magnitude range from 1.4 to 8.0. Since 1994, MLR (Muntele Rosu) station has become part of the GEOFON network and was provided with high performance broad band instruments. At Bucharest and Timisoara data centers, an automated and networked seismological system performs the on-line digital acquisition and processing of the telemetered data. Automatic processing includes discrimination between local and distant seismic events, earthquake location and magnitude computation, and source parameter determination for local earthquakes. The results are rapidly distributed via Internet, to several seismological services in Europe and USA, to be used in the association/confirmation procedures. Plans for new developments of the network include the upgrade from analog to digital telemetry and new stations for monitoring local seismicity. (authors)

A personal computer code for seismic evaluations of nuclear power plants facilities

International Nuclear Information System (INIS)

Xu, J.; Philippacopoulos, A.J.; Graves, H.

1990-01-01

The program CARES (Computer Analysis for Rapid Evaluation of Structures) is an integrated computational system being developed by Brookhaven National Laboratory (BNL) for the U.S. Nuclear Regulatory Commission. It is specifically designed to be a personal computer (PC) operated package which may be used to determine the validity and accuracy of analysis methodologies used for structural safety evaluations of nuclear power plants. CARES is structured in a modular format. Each module performs a specific type of analysis i.e., static or dynamic, linear or nonlinear, etc. This paper describes the various features which have been implemented into the Seismic Module of CARES

Pre-Operational Seismic Walk-Through of NPPs in India

International Nuclear Information System (INIS)

Soni, R.S.; Mishra, R.K.; Agrawal, M.K.; Reddy, G.R.; Kushwaha, H.S.; Venkat Raj, V.; Badrinarayan, G.; Hawaldar, R.V.; Ingole, S.M.

2002-01-01

In nuclear power plants, it is essential to design the various safety and safety related systems and components of the plant in such a manner that they maintain their structural integrity as well as serve their functional performance during a seismic event. The pre-operational seismic walk-through helps in ensuring the installation of various seismic supports as per design intent, identifying the areas where supports are inadequate, identifying the interaction concerns between the systems of various safety classes and locating the various undesired loose, untied / unanchored components, tools, etc. used during the construction activity. A detailed procedure for the pre-operational seismic walk-through of the NPPs was therefore, prepared. Since the types and locations of seismic supports for the various systems and components of the plant had been already reviewed, the major emphasis during the walk-through was laid on their proper installation. (authors)

Signal-to-noise ratio application to seismic marker analysis and fracture detection

Science.gov (United States)

Xu, Hui-Qun; Gui, Zhi-Xian

2014-03-01

Seismic data with high signal-to-noise ratios (SNRs) are useful in reservoir exploration. To obtain high SNR seismic data, significant effort is required to achieve noise attenuation in seismic data processing, which is costly in materials, and human and financial resources. We introduce a method for improving the SNR of seismic data. The SNR is calculated by using the frequency domain method. Furthermore, we optimize and discuss the critical parameters and calculation procedure. We applied the proposed method on real data and found that the SNR is high in the seismic marker and low in the fracture zone. Consequently, this can be used to extract detailed information about fracture zones that are inferred by structural analysis but not observed in conventional seismic data.

Lessons learned from NRC systematic evaluation program seismic review

International Nuclear Information System (INIS)

Cheng, T.M.; Hermann, R.A.; Russell, W.T.

1983-01-01

In October 1977, the Nuclear Regulatory Commission approved initiation of Phase II of the Systematic Evaluation Program (SEP) which consists of a plant-specific reassessment of the safety of 11 older operating nuclear reactors. Many safety criteria have rapidly evolved since the time of initial licensing of these plants. The purpose of the SEP is to develop a current documented basis for the safety of these older facilities by comparing them to current criteria. Phase I of the SEP developed a comprehensive list of 137 topics of safety significance which collectively affect the plant's capability to respond to various Design Basis Events (DBEs). Seismic Design Consideration is one of the 137 safety topics. (orig./GL)

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)

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

Calibration of Seismic Attributes for Reservoir Characterization; ANNUAL

International Nuclear Information System (INIS)

Pennington, Wayne D.; Acevedo, Horacio; Green, Aaron; Len, Shawn; Minavea, Anastasia; Wood, James; Xie, Deyi

2002-01-01

This project has completed the initially scheduled third year of the contract, and is beginning a fourth year, designed to expand upon the tech transfer aspects of the project. From the Stratton data set, demonstrated that an apparent correlation between attributes derived along 'phantom' horizons are artifacts of isopach changes; only if the interpreter understands that the interpretation is based on this correlation with bed thickening or thinning, can reliable interpretations of channel horizons and facies be made. From the Boonsville data set , developed techniques to use conventional seismic attributes, including seismic facies generated under various neural network procedures, to subdivide regional facies determined from logs into productive and non-productive subfacies, and developed a method involving cross-correlation of seismic waveforms to provide a reliable map of the various facies present in the area. The Teal South data set provided a surprising set of data, leading us to develop a pressure-dependent velocity relationship and to conclude that nearby reservoirs are undergoing a pressure drop in response to the production of the main reservoir, implying that oil is being lost through their spill points, never to be produced. The Wamsutter data set led to the use of unconventional attributes including lateral incoherence and horizon-dependent impedance variations to indicate regions of former sand bars and current high pressure, respectively, and to evaluation of various upscaling routines

Local seismic network at the Olkiluoto site. Annual report 2002-2004

Energy Technology Data Exchange (ETDEWEB)

Saari, J. [Enprima Oy, Vantaa (Finland)

2005-09-15

In Olkiluoto, Posiva Oy has operated a local seismic network since February 2002. In the beginning, the network consisted of six seismic stations. Later, in June 2004, the seismic network was expanded with two new seismic stations. At that time started the excavation of the underground characterisation facility (the ONKALO) and the basic operation procedure was changed more suitable for the demands of the new situation. The purpose of the microearthquake measurements at Olkiluoto is to improve understanding of the structure, behaviour and long term stability of the bedrock. The studies include both tectonic and excavation-induced microearthquakes. An additional task of monitoring is related to safeguarding of the ONKALO. This report gives the results of microseismic monitoring during the years 2002 - 2004. Also the changes in the structure and the operation procedure of the network are described. The network has operated nearly continuously. The longest interruption occurred 16.-17.6.2004, when two new seismic stations were installed in the network and the operation procedure was changed. Altogether 757 events have been located in the Olkiluoto area. The magnitudes of the observed events range from ML = -3.5 to ML = 1.2. All of them are explosions or other artificial events. So far, none of the 757 observed events can be classified as microearthquakes. Five of the events have characteristics that make the origin of the recorded signal uncertain. They are quite unlikely microearthquakes, but they are not typical examples of artificial seismic signals either. When the experience and the data set of the Olkiluoto microearthquakes increase the identification of events will be more definite. Evidence of activity that would has influence on the safety of the ONKALO, have not found. (orig.)

Local seismic network at the Olkiluoto site. Annual report 2002-2004

International Nuclear Information System (INIS)

Saari, J.

2005-09-01

In Olkiluoto, Posiva Oy has operated a local seismic network since February 2002. In the beginning, the network consisted of six seismic stations. Later, in June 2004, the seismic network was expanded with two new seismic stations. At that time started the excavation of the underground characterisation facility (the ONKALO) and the basic operation procedure was changed more suitable for the demands of the new situation. The purpose of the microearthquake measurements at Olkiluoto is to improve understanding of the structure, behaviour and long term stability of the bedrock. The studies include both tectonic and excavation-induced microearthquakes. An additional task of monitoring is related to safeguarding of the ONKALO. This report gives the results of microseismic monitoring during the years 2002 - 2004. Also the changes in the structure and the operation procedure of the network are described. The network has operated nearly continuously. The longest interruption occurred 16.-17.6.2004, when two new seismic stations were installed in the network and the operation procedure was changed. Altogether 757 events have been located in the Olkiluoto area. The magnitudes of the observed events range from ML = -3.5 to ML = 1.2. All of them are explosions or other artificial events. So far, none of the 757 observed events can be classified as microearthquakes. Five of the events have characteristics that make the origin of the recorded signal uncertain. They are quite unlikely microearthquakes, but they are not typical examples of artificial seismic signals either. When the experience and the data set of the Olkiluoto microearthquakes increase the identification of events will be more definite. Evidence of activity that would has influence on the safety of the ONKALO, have not found. (orig.)

Component fragility analysis methodology for seismic risk assessment projects. Proven PSA safety document processing and assessment procedures

International Nuclear Information System (INIS)

Kolar, Ladislav

2013-03-01

The seismic risk task assessment task should be structured as follows: (i) Define all reactor unit building structures, components and equipment involved in the creation of an initiating event (IE) induced by an seismic event or contributing to the reliability of reactor unit response to an IE; (ii) construct and estimate of the fragility curves for the building and component groups sub (i); (iii) determine the HCLPF for each group of buildings, components or equipment; (iv) determine the nuclear source's seismic resistance (SME) as the minimum HCLPF from the group of equipment in the risk-dominant scenarios; (v) define the risk-limiting group of components, equipment and building structures to the SME value; (vi) based on the fragility levels, identify component groups for which a more detailed fragility analysis is needed; and (vii) recommend groups of equipment or building structures that should be taken into account with respect to the seismic risk, i.e. such groups of equipment or building structures as exhibit a low seismic resistance (HCLPF) and, at the same time, are involved to a significant extent in the reactor unit's seismic risk (are present in the dominant risk scenarios). (P.A.)

Annual Hanford seismic report - fiscal year 1996

International Nuclear Information System (INIS)

Hartshorn, D.C.; Reidel, S.P.

1996-12-01

Seismic monitoring (SM) at the Hanford Site was established in 1969 by the US Geological Survey (USGS) under a contract with the US Atomic Energy Commission. Since 1980, the program has been managed by several contractors under the US Department of Energy (USDOE). Effective October 1, 1996, the Seismic Monitoring workscope, personnel, and associated contracts were transferred to the USDOE Pacific Northwest National Laboratory (PNNL). SM is tasked to provide an uninterrupted collection and archives of high-quality raw and processed seismic data from the Hanford Seismic Network (HSN) located on and encircling the Hanford Site. SM is also tasked to locate and identify sources of seismic activity and monitor changes in the historical pattern of seismic activity at the Hanford Site. The data compiled are used by SM, Waste Management, and engineering activities at the Hanford Site to evaluate seismic hazards and seismic design for the Site

Pickering seismic safety margin

International Nuclear Information System (INIS)

Ghobarah, A.; Heidebrecht, A.C.; Tso, W.K.

1992-06-01

A study was conducted to recommend a methodology for the seismic safety margin review of existing Canadian CANDU nuclear generating stations such as Pickering A. The purpose of the seismic safety margin review is to determine whether the nuclear plant has sufficient seismic safety margin over its design basis to assure plant safety. In this review process, it is possible to identify the weak links which might limit the seismic performance of critical structures, systems and components. The proposed methodology is a modification the EPRI (Electric Power Research Institute) approach. The methodology includes: the characterization of the site margin earthquake, the definition of the performance criteria for the elements of a success path, and the determination of the seismic withstand capacity. It is proposed that the margin earthquake be established on the basis of using historical records and the regional seismo-tectonic and site specific evaluations. The ability of the components and systems to withstand the margin earthquake is determined by database comparisons, inspection, analysis or testing. An implementation plan for the application of the methodology to the Pickering A NGS is prepared

Evaluation of Procedures for Backcalculation of Airfield Pavement Moduli

Science.gov (United States)

2015-08-01

ER D C/ G SL T R -1 5 -3 1 Evaluation of Procedures for Backcalculation of Airfield Pavement Moduli G eo te ch n ic al a n d S tr u...August 2015 Evaluation of Procedures for Backcalculation of Airfield Pavement Moduli Lucy P. Priddy and Carlos R. Gonzalez Geotechnical and...USAF’s) airfield pavement structural evaluation procedures. Determining the structural integrity of airfield pavement relies on the analysis of

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)

Demonstration of NonLinear Seismic Soil Structure Interaction and Applicability to New System Fragility Seismic Curves

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology

2014-09-01

Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in general this approach has been conservative, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). SPRAs are performed by convolving the seismic hazard (the frequency of certain magnitude events) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, fragility or capacity of structures, systems and components (SSC), and systems analysis. Figure 1 provides a high level overview of the risk quantification process. The focus of this research is on understanding and removing conservatism (when possible) in the quantification of seismic risk at NPPs.

Nuclear power plants seismic review programme in Spain

International Nuclear Information System (INIS)

Sanchez Cabanero, J.G.; Jimenez Juan, A.

1995-01-01

This presentation deals with seismic design and seismic reevaluation of Spanish operating nuclear power plants. The Spanish NPPs owners Probabilistic seismic hazard study requires an independent and deep review of methodology, modelling techniques, data management and method of eliciting in order to make decision on its acceptability. It reflects the opinion of only one expert tem and it would be necessary to involve more expert opinions to consider the uncertainties. It is proposed to evaluate the probabilistic seismic hazard study and the seismic categorisation

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

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)

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.)

IEEE C37.98-1987: IEEE standard seismic testing of relays

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard specifies the procedures to be used in the seismic testing of relays used in power system facilities. The standard is concerned with the determination of the seismic fragility level of relays and also gives recommendations for proof testing. The purpose of this standard is to establish procedures for determining the seismic capabilities of protective and auxiliary relays. These procedures employ what has been called fragility testing in IEEE Std 344-1987. To define the conditions for fragility testing of relays, parameters in three separate areas must be specified. In general, they are (1) the electrical settings and inputs to the relay, and other information to define its conditions during the test; (2) the change in state, deviation in operating characteristics or tolerances, or other change of performance of the relay that constitutes failure; (3) the seismic vibration environment to be imposed during the test. Since it is not possible to define the conditions for every conceivable application for all relays, those parameters, which in practice encompass the majority of applications, have been specified in this standard. When the application of the relay is other than as specified under any of (1), (2), and (3), or if it is not practical to apply existing results of fragility tests to that new application, then proof testing must be performed for that new case

Development of analysis methods for seismically isolated nuclear structures

International Nuclear Information System (INIS)

Yoo, Bong; Lee, Jae-Han; Koo, Gyeng-Hoi

2002-01-01

KAERI's contributions to the project entitled Development of Analysis Methods for Seismically Isolated Nuclear Structures under IAEA CRP of the intercomparison of analysis methods for predicting the behaviour of seismically isolated nuclear structures during 1996-1999 in effort to develop the numerical analysis methods and to compare the analysis results with the benchmark test results of seismic isolation bearings and isolated nuclear structures provided by participating countries are briefly described. Certain progress in the analysis procedures for isolation bearings and isolated nuclear structures has been made throughout the IAEA CRPs and the analysis methods developed can be improved for future nuclear facility applications. (author)

Evaluation of near-surface attenuation of S-waves based on PS logging and vertical array seismic observation

International Nuclear Information System (INIS)

Kobayashi, Genyu

2014-01-01

As a result of the lessons learned from the experience of Kashiwazaki-Kariwa NPP due to the 2007 Niigata Chuetsu Oki Earthquake, it has become clear that a rational method of near-surface attenuation characteristics covering a depth range from engineering bedrock to seismic bedrock urgently needs to be established. JNES performed PS logging and vertical array seismic ground motion observation at a soil ground site (SODB 1. site), sedimentary rock site, and an igneous rock site (SODB 2. site), and proposed an evaluation method of attenuation characteristics (site characteristics) for the deep underground. (author)

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

International Nuclear Information System (INIS)

Payne, Suzette; Coppersmith, Ryan; Coppersmith, Kevin; Rodriguez-Marek, Adrian; Falero, Valentina Montaldo; Youngs, Robert

2016-01-01

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Naval Reactors Facility (NRF), and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) (Figure 1-1). The PSHA followed the approaches and procedures appropriate for a Study Level 1 provided in the guidance advanced by the Senior Seismic Hazard Analysis Committee (SSHAC) in U.S. Nuclear Regulatory Commission (NRC) NUREG/CR-6372 and NUREG-2117 (NRC, 1997; 2012a). The SSHAC Level 1 PSHAs for MFC and ATR were conducted as part of the Seismic Risk Assessment (SRA) project (INL Project number 31287) to develop and apply a new-risk informed methodology, respectively. The SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels. The SRA project is developing a new risk-informed methodology that will provide a systematic approach for evaluating the need for an update of an existing PSHA. The new methodology proposes criteria to be employed at specific analysis, decision, or comparison points in its evaluation process. The first four of seven criteria address changes in inputs and results of the PSHA and are given in U.S. Department of Energy (DOE) Standard, DOE-STD-1020-2012 (DOE, 2012a) and American National Standards Institute/American Nuclear Society (ANSI/ANS) 2.29 (ANS, 2008a). The last three criteria address evaluation of quantitative hazard and risk-focused information of an existing nuclear facility. The seven criteria and decision points are applied to Seismic Design Category (SDC) 3, 4, and 5, which are defined in American Society of Civil Engineers/Structural Engineers Institute (ASCE/SEI) 43-05 (ASCE, 2005). The application of the criteria and decision points could lead to an update or could determine that such update is not necessary.

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Payne, Suzette [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coppersmith, Ryan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coppersmith, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rodriguez-Marek, Adrian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Falero, Valentina Montaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngs, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Naval Reactors Facility (NRF), and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) (Figure 1-1). The PSHA followed the approaches and procedures appropriate for a Study Level 1 provided in the guidance advanced by the Senior Seismic Hazard Analysis Committee (SSHAC) in U.S. Nuclear Regulatory Commission (NRC) NUREG/CR-6372 and NUREG-2117 (NRC, 1997; 2012a). The SSHAC Level 1 PSHAs for MFC and ATR were conducted as part of the Seismic Risk Assessment (SRA) project (INL Project number 31287) to develop and apply a new-risk informed methodology, respectively. The SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels. The SRA project is developing a new risk-informed methodology that will provide a systematic approach for evaluating the need for an update of an existing PSHA. The new methodology proposes criteria to be employed at specific analysis, decision, or comparison points in its evaluation process. The first four of seven criteria address changes in inputs and results of the PSHA and are given in U.S. Department of Energy (DOE) Standard, DOE-STD-1020-2012 (DOE, 2012a) and American National Standards Institute/American Nuclear Society (ANSI/ANS) 2.29 (ANS, 2008a). The last three criteria address evaluation of quantitative hazard and risk-focused information of an existing nuclear facility. The seven criteria and decision points are applied to Seismic Design Category (SDC) 3, 4, and 5, which are defined in American Society of Civil Engineers/Structural Engineers Institute (ASCE/SEI) 43-05 (ASCE, 2005). The application of the criteria and decision points could lead to an update or could determine that such update is not necessary.

A personal computer code for seismic evaluations of nuclear power plant facilities

International Nuclear Information System (INIS)