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Sample records for npp seismic design

  1. 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)

  2. 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)

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

    International Nuclear Information System (INIS)

    Kaliberda, Y.V.

    2003-01-01

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

  4. Seismic analysis and design of NPP structures

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. The seismic reassessment Mochovce NPP

    International Nuclear Information System (INIS)

    Baumeister, P.

    2004-01-01

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

  7. Seismic characterization of the NPP Krsko site

    International Nuclear Information System (INIS)

    Obreza, J.

    2000-01-01

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

  8. 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

  9. Seismic safety of building structures of NPP Kozloduy III

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

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

  11. Seismic 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

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

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

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

  13. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  14. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-15

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

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

    International Nuclear Information System (INIS)

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

    2005-03-01

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

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

    International Nuclear Information System (INIS)

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

    1992-11-01

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

  18. 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.)

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

    International Nuclear Information System (INIS)

    Bojadziev, Z.; Kostov, M.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. Seismic hazard analysis of the NPP Kozloduy site

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  2. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  4. 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.

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

    International Nuclear Information System (INIS)

    Zadoyan, P.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  7. 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.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  12. 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)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  14. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

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

    International Nuclear Information System (INIS)

    Birbraer, A.N.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-10-01

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

  18. 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)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    International Nuclear Information System (INIS)

    Yakovlev, E.A.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  3. Seismic design and qualification for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

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

  4. 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)

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

    International Nuclear Information System (INIS)

    Rieck, P.J.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1999-01-01

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

  7. Seismic design considerations for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Chang Xiangdong; Zhou Bengang; Zhao Lianda

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Prachar, Ivan; Vacek, Jiri; Heralecky, Pavel

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  11. Determination of the NPP Cernavoda reactor building seismic response

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Muzzi, F.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  12. Constructive approaches to the space NPP designing

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  13. Seismic design features of the ACR Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Belene NPP

    International Nuclear Information System (INIS)

    Tsvetanov, P.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  16. 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.

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

    International Nuclear Information System (INIS)

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Viallet, Emmanuel

    2014-01-01

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

  1. 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

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

    International Nuclear Information System (INIS)

    Stoyanov, G.; Jordanov, M.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Sachanski, S.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-01

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

  6. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Hauptenbuchner, B.; David, M.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  9. New trends in designing NPP control boards

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

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

  11. 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.)

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

    International Nuclear Information System (INIS)

    Boyadjiev, Z.

    1995-01-01

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

  13. 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)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Stojka, Tibor

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

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

  18. 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)

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

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    2003-01-01

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

  20. 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.

  1. 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

  2. Structure study and design of Qinshan NPP PCCV

    International Nuclear Information System (INIS)

    Xia Zufeng; Xu Yongzhi; Wang Tianzhen; Wu Jibiao

    1993-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  4. Role of PRA in new NPP projects

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Kalincik, L.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Hirose, Jiro; Muramatsu, Ken

    2002-03-01

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

  8. Isolation systems influence in the seismic loading propagation analysis applied to an innovative near term reactor

    International Nuclear Information System (INIS)

    Lo Frano, R.; Forasassi, G.

    2010-01-01

    Integrity of a Nuclear Power Plant (NPP) must be ensured during the plant life in any design condition and, particularly, in the event of a severe earthquake. To investigate the seismic resistance capability of as-built structures systems and components, in the event of a Safe Shutdown Earthquake (SSE), and analyse its related effects on a near term deployment reactor and its internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was applied considering, also, the use of innovative anti-seismic techniques. In this paper the attention is focused on the use and influence of seismic isolation technologies (e.g. isolators based on passive energy dissipation) that seem able to ensure the full integrity and operability of NPP structures, to enhance the seismic safety (improving the design of new NPPs and if possible, to retrofit existing facilities) and to attain a standardization plant design. To the purpose of this study a numerical assessment of dynamic response/behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) are herein presented and compared in order to highlight the isolation technique effectiveness.

  9. Seismic Data Gathering and Validation

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    To increase the safety of a NPP located on a seismic site, the seismic acceleration level to which the NPP should be qualified must be as representative as possible for that site, with a conservative degree of safety but not too exaggerated. The consideration of the seismic events affecting the site as independent events and the use of statistic methods to define some safety levels with very low annual occurrence probability (10 -4 ) may lead to some exaggerations of the seismic safety level. The use of some very high value for the seismic acceleration imposed by the seismic safety levels required by the hazard analysis may lead to very costly technical solutions that can make the plant operation more difficult and increase maintenance costs. The considerations of seismic events as a time series with dependence among the events produced, may lead to a more representative assessment of a NPP site seismic activity and consequently to a prognosis on the seismic level values to which the NPP would be ensured throughout its life-span. That prognosis should consider the actual seismic activity (including small earthquakes in real time) of the focuses that affect the plant site. The paper proposes the applications of Autoregressive Time Series to issue a prognosis on the seismic activity of a focus and presents the analysis on Vrancea focus that affects NPP Cernavoda site, by this method. The paper also presents the manner to analyse the focus activity as per the new approach and it assesses the maximum seismic acceleration that may affect NPP Cernavoda throughout its life-span (∼ 30 years). Development and applications of new mathematical analysis method, both for long - and short - time intervals, may lead to important contributions in the process of foretelling the seismic events in the future. (authors)

  11. Review of the main issues connected with the Belene NPP project

    International Nuclear Information System (INIS)

    Simeonov, P.

    1996-01-01

    The basic seismic studies connected with the construction of the second NPP in Bulgaria are outlined. The site selected for this plant is located on the right bank of the Donau River, near the Belene town. The seismicity of the site is investigated by the Bulgarian Academy of Sciences (BAS) and Energoproject Ltd., Sofia, and assessed by IAEA, Siemens, EQE and Westinghouse experts. The suggested coefficient for maximum seismic acceleration of 0.24 g takes into account the safety of the site, structures and equipment of the NPP. The conclusion made is that from seismic point of view the site is not hazardous. Construction works started in 1987 and were stopped in 1990 due to financial and socio-ecological reasons. The author's opinion is that the construction of the Belene NPP has no alternative and should be completed providing higher safety and reliability of the plant operation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    International Nuclear Information System (INIS)

    David, M.

    1997-01-01

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

  14. 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)

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

    International Nuclear Information System (INIS)

    Kusnowo, A.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Science.gov (United States)

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

    2015-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

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

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

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Proskuryakov, Konstantin

    2011-01-01

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

  4. Seismic methodology in determining basis earthquake for nuclear installation

    International Nuclear Information System (INIS)

    Ameli Zamani, Sh.

    2008-01-01

    Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

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

    International Nuclear Information System (INIS)

    Hintergraeber, M.; Wittmann, R.

    1985-01-01

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

  6. Risk based seismic design criteria

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Kennedy, R.P.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Freeland, Kent

    2013-01-01

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

  9. Geological-Hydrological Site Evaluation for NPP Planning

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  10. Geological-Hydrological Site Evaluation for NPP Planning

    International Nuclear Information System (INIS)

    Faust, Brigitte; Mini, Paolo

    2008-01-01

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

  11. Sitting Safety Aspects of Second Romanian NPP

    International Nuclear Information System (INIS)

    Mauna, T.

    2010-01-01

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

  12. Seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    1996-01-01

    This report covers the review of the already completed studies, namely, safe shutdown system identification and classification for Bohunice NPP and the comparative study of standards and criteria. It contains a report on currently ongoing studies concerning seismic margin assessment and earthquake experience based methods in application for seismic evaluation and verification of structures and equipment components of the operating WWER-440/213 type NPPs. This is based on experiences obtained from Paks NPP. The work plan for the remaining period of Benchmark CRP and the new proposals are included. These are concerned with seismic evaluation of selected safety related mechanical equipment and pipes of Paks NPP, and the actual seismic issues of the Temelin WWER-1000 type NPP

  13. Research on performance-based seismic design criteria

    Institute of Scientific and Technical Information of China (English)

    谢礼立; 马玉宏

    2002-01-01

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

  14. Core design methodology and software for Temelin NPP

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  16. The Role of Tectonic and Seismicity in Siting of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Abdel Aziz, M.A.H.

    2008-01-01

    The site selection for the a nuclear power plant (NPP) is controlled by many criteria. One of the most important criterion is the tectonic and seismicity of the site and its surroundings. Since, it is preferable the site in concern is characterized by low tectonic and low seismicity to avoid the damage effects associated with the occurrence of destructive earthquakes. The investigation of the tectonic and seismicity maps of egypt has been carried out to candidate potential areas or sites for nuclear power plant installation from seismicity point of view. Also, the design basis ground motion in terms of peak ground acceleration and response spectra of some of the potential sites are defined through the conduct of probabilistic seismic hazard analysis The study revealed that although there is no criterion to exclude areas of high tectonic and high seismicity as potential sites for nuclear power plant installation but, it is preferable avoiding such areas. This is attributed to the critical seismic curve that characterizes such areas and is required high seismic design levels to resist the destructive vibratory ground motion associated with the expected earthquake. Consequently, the required high seismic design levels will have a negative impact on the economic cost of the facility compared with that built in low and moderate seismic areas. Hence, areas like the gulf of suez, the northern part of the Red Sea and the southern part of Sinai Peninsula should be avoided as potential sites for NPP from the tectonic and seismicity point of view. On the other hand, areas like Nile delta and its valley, the Northern and Southern parts of Western desert and the central and southern parts of the Eastern Desert should be candidate as potential sites on condition, the other criteria meet the IAEA's regulations. Also, the seismic hazard curve of the Northwest littoral zone reflects low design basis ground motion values compared with the Nile delta region

  17. Collection and accumulation of seismic safety research findings, and considerations for information dissemination

    International Nuclear Information System (INIS)

    2013-01-01

    Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

  18. Collection and accumulation of seismic safety research findings, and considerations for information dissemination

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Seismic Safety Division of JNES is collecting and analyzing the findings of seismic safety research, and is developing a system to organize and disseminate the information internally and internationally. These tasks have been conducted in response to the lessons learned from Fukushima Daiichi NPP accident. The overview of the tasks is as follows; 1) Collection of the knowledge and findings from seismic safety research. JNES collects information on seismic safety researches including the 2011 off the Pacific coast of Tohoku Earthquake. The information is analyzed whether it is important for regulation to increase seismic safety of NPP. 2) Constructing database of seismic safety research. JNES collects information based on documents published by committee and constructs database of active faults around NPP sites in order to incorporate in the seismic safety review. 3) Dissemination of information related to seismic safety. JNES disseminates outcomes of own researches internally and internationally. (author)

  19. 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

  20. Assessing seismic adequacy of existing nuclear power plant structures

    International Nuclear Information System (INIS)

    Belyaev, V.; Vinogradov, V.; Privalov, S.; Shishenin, V.

    2003-01-01

    Nowadays Russia's specialists perform a huge amount of works to revaluate the NPP safety. These works are certain to include refinement of NPP safety assessment under the effects of specific dynamic loads, earthquake effects included. It should be noted, that a number of Russian NPPs now in operation had been designed either with no account of these loads, or under the requirements which are underestimated as compared with the modern requirements on the external load composition and rate. Revaluation of NPP seismic safety is based on the results of the works taken under orderly sequence on assessment of (1) seismic input and ground effects; (2) structure response and state; (3) equipment and pipelines response and state. The paper considers the methods of NPP structures response and state assessment. Therewith we assume that ground motion predicted behavior at the construction basement has been preset for the SSE and OBE conditions and the effects of soil-structure interaction, including the situation of possible soft soil liquefaction. Necessity to determine both the reaction of a construction and its state as a whole as well as its elements reaction, to evaluate their bearing capacity and destruction zones formation makes it necessary to make up a detailed structural model, which is usually a finite element one. Since seismic revaluation is to be performed for the existing structures, characteristics of which can substantially differ from the design ones, revealing the actual state of this structures becomes critical. If the real values of physical and mechanical properties of the structure materials, connections of elements etc. are used as initial data in a structural model this permits to increase the design assessment credibility and reliability substantially. The paper analyzes the results of determining these initial assessments while inspecting several Russian NPPs on the basis of a 'combined' method. This method is realized at two consecutive stages. The

  1. Seismic analysis of steam generator and parameter sensitivity studies

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  2. 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

  3. Structural response of Paks NPP WWER-440 MW main building complex to blast input motion. Final report

    International Nuclear Information System (INIS)

    1999-01-01

    The Soviet standard design units WWER-440/213 type installed in Paks NPP were not originally designed for a Safe Shutdown Earthquake. At the time of selection of Paks site on the basis of historical earthquake data was supposed that the maximum earthquake is of grade V according MSK-64 scale. This seismicity level had not required any special measures to account for seismic event effects on the Main Building Complex Structure. Current site seismicity studies reveal that the seismic hazard for the site significantly exceeds the originally estimated. In addition the safety rules and seismic code requirements became more rugged. As a part of the activities to increase the seismic safety of the Paks NPP the study on dynamic behaviour of the Main Building Complex Structure has been performed with support of IAEA. The explosion full scale tests were carried out for determining the dynamic behaviour of the structure and for assessment of the Soil Structure Interaction (SSI) effects in the modelling and analysis procedures, used in the dynamic response analyses. The objective of the project was to evaluate the blast response of the WWER-440/213 Main Building Complex at Paks NPP, based on the data available for the soil properties, recorded free-field blast input motion, and structural design. The scope of EQE-Bulgaria study was to conduct a state-of-the-art SSI analysis with a multiple foundations supported model of the Main Building Complex to assess the structure blast response. The analysis was focused on a modelling technique that assess realistically the SSI effects on the dynamic response of a structure supported on multiple foundation instead of simplified, but more conservative techniques. The scope of research was covered splitting the study into the following steps: development of a twin units model for Main Building Complex structure; development of a Low Strain Soil Properties Model; development of SSI Parameters consisting of a Multiple Foundations System

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  6. Latest standards on seismic resistance related to research activities

    International Nuclear Information System (INIS)

    Juhasova, Emilia

    2002-01-01

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

  7. Seismic soil–structure interaction analysis of a nuclear power plant building founded on soil and in degraded concrete stiffness condition

    International Nuclear Information System (INIS)

    Farahani, Reza V.; Dessalegn, Tewodros M.; Vaidya, Nishikant R.; Bazan-Zurita, Enrique

    2016-01-01

    . Therefore, it is important to consider SSI parameter variability in the seismic design or evaluation of NPP buildings that bear on soil.

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

    International Nuclear Information System (INIS)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  10. 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

  11. Seismic evaluation of existing nuclear power plants

    International Nuclear Information System (INIS)

    2003-01-01

    programmes at existing operating plants are plant specific or regulatory specific. This means that this report is meant to define the minimum generic requirements and may need to be supplemented on a plant specific basis to consider particular aspects of the original design basis. Among the options available, two methods are particularly appropriate for assessing the seismic safety of facilities, the seismic margin assessment (SMA) method and the seismic probabilistic safety assessment (SPSA) method. Both SMA and SPSA are discussed in this report. Current NPP design criteria and comprehensive seismic design procedures, as applied to the design of new facilities but using a re-evaluated seismic input, may be applied in the seismic evaluation programme. It is noted that these would be a conservative and usually expensive approach for evaluation of an existing operating facility and they are not discussed further in this report. Evaluation of existing NPPs may result in the identification of items of the SSSC list which have to be upgraded. Upgrading itself is not covered by this Safety Report; however, some general principles are presented in order to preserve consistency between evaluation and upgrading processes. (It should be pointed out that when an upgrading programme has to be carried out, it necessitates more engineering resources than the evaluation process does; similarly upgrading is too large and complex a matter to be covered by this Safety Report.) Section 2 presents the general philosophy of seismic evaluation; Section 3 discusses data collection and investigations; Section 4 is devoted to seismic hazard assessment; Section 5 discusses the safety analysis of the NPP; Section 6 discusses the practice of walkdown; Section 7 covers the criteria and methods used for seismic capacity assessment of SSSCs; Section 8 discusses the principle of the design of a possible seismic upgrading; Section 9 specifies some rules of quality assurance and organization

  12. NRC Seismic Design Margins Program Plan

    International Nuclear Information System (INIS)

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

    1985-08-01

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

  13. Study on integrated design and analysis platform of NPP

    International Nuclear Information System (INIS)

    Lu Dongsen; Gao Zuying; Zhou Zhiwei

    2001-01-01

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

  14. Position paper: Seismic design criteria

    International Nuclear Information System (INIS)

    Farnworth, S.K.

    1995-01-01

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

  15. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

  16. Determination of Design Basis Earthquake ground motion

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  17. Determination of Design Basis Earthquake ground motion

    International Nuclear Information System (INIS)

    Kato, Muneaki

    1997-01-01

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

  18. 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

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

    International Nuclear Information System (INIS)

    Ivkov, Igor

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Freiman, M.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Jing Xu

    2012-01-01

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

  2. Analysis and evaluation of seismic response of reactor building for Daya Bay Nuclear Power Plant

    International Nuclear Information System (INIS)

    Li Zhongcheng; China Guangdong Nuclear Power Company, Shenzhen; Li Zhongxian

    2005-01-01

    Daya Bay NPP has been operating safely and stably over 10 years since 1994, and its' seismic analysis of nuclear island was in accordance with the approaches in RCC-G standard for the model M310, in which the Simplified Impedance Matrix Method (SIMM) 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 Daya Bay NPP, in his 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 is hopefully to provide some important references to the assessment of seismic safety margin for the operating NPPs. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Tripathi, B.P.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    1987-12-01

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

  7. 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

  8. 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.

  9. 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

  10. Seismic design of reactors in NUCEF

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  11. Experimental investigation on full scale RC beam-column joint of NPP structures

    International Nuclear Information System (INIS)

    Thandavamoorthy, T.S.; Lakshmanan, N.; Reddy, G.R.; Kushwaha, H.S.

    2003-01-01

    The Nuclear Power Plant (NPP) structures in India are constructed using reinforced concrete. The beam-column joint in these structures are critical sub-assemblages because they ensure continuity of a structure and transfer forces from one element to another. Under seismic excitation, the beam-column joint region is subjected to horizontal and vertical shear forces whose magnitudes are typically many times higher than those within the adjacent beams and columns. In view of the increased incidence of seismicity in the country, the safety of these structures against earthquake loading assumes greater significance. There is a growing need to look into the seismic safety aspect of existing RC frame type structures in NPPs, which have been designed as per codes prevalent at the time of their construction. Seismic performance of such joints has not been studied extensively in India. Therefore experimental testing of full scale joint identical to those available in the existing NPP structures, was carried out to study its behaviour and evaluate its capacity. The size of the beam of the joint was 2000 mm x 610 mm x 915 mm and column 2915 mm x 610 mm x 915 mm. The percentage reinforcement of the beam was 4.95 and column 1.5. Such full scale and heavily reinforced concrete joint was cast successfully in the laboratory and tested under monotonic loading. The paper presents a complete description of the experimental testing, observations made during testing as for cracking, deflection and rotation of joint, discussion of results obtained, etc. Conclusions drawn from the investigation are also presented. (author)

  12. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  13. Effects of Moat Wall Impact on the Seismic Response of Base Isolated Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Kim, Jung Han; Mosqueda, Gilberto; Sarebanhab, Alireza

    2015-01-01

    Implementation of the moat wall as a hard stop for isolated NPP is able to decrease the displacement demand. However, since the NPP is relatively heavy structure designed to remain elastic during a seismic event, the moat wall and backfill soil flexibility result in a significant penetration into the wall. As a result, isolators undergo displacements exceeding design considerations based on the clearance to hard stop. Isolators should be tested beyond the hard stop to ensure functionality in the case of an extreme earthquake

  14. Probabilistic site dependent design spectra for a NPP

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. Design experience on seismically isolated buildings

    International Nuclear Information System (INIS)

    Giuliani, G.C.

    1989-01-01

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

  16. 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.

  17. Engineering safety review mission Krsko NPP external events PSA. Ljubljana, Slovenia 19-23 February 1996. Final report

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Smith, P.

    1996-01-01

    Within the scope of the TC Project RER/9/035, a review mission visited Ljubljana, Slovenia, 19-23 February 1996. Two outside experts, Messrs. R.J. Budnitz (USA) and Paul Smith (USA), as well as a staff member, A. Guerpinar (ESS-NSNI) took part in the review. The purpose of the mission was to assist the Slovenian Nuclear Safety Administration to review the external events PSA prepared by Krsko NPP consultants Westinghouse Energy Systems Europe and EQE International. Another seismic safety review was performed concurrently in Ljubljana involving the investigations in relation to the tectonic stability and reassessment of the design basis ground motion characterization for the Krsko NPP site

  18. Harmonizing seismic hazard assessments for nuclear power plants

    International Nuclear Information System (INIS)

    Mallard, D.J.

    1993-01-01

    Even a cursory comparison between maps of global seismicity and NPP earthquake design levels reveals many inconsistencies. While, in part, this situation reflects the evolution in understanding of seismic hazards, mismatches can also be due to ongoing differences in the way the hazards are assessed and in local regulatory requirements. So far, formal international consensus has only been able to encompass broad principles, such as those recently recommended by the International Atomic Energy Agency, and even these can raise many technical issues, particularly relating to zones of diffuse seismicity. In the future, greater harmonisation in hazard assessments and, to some extent, in earthquake design levels could emerge through the more widespread use of probabilistic methods. International collaborative ventures and joint projects will be important for resolving anomalies in the existing databases and their interpretations, and for acquiring new data, but to achieve their ideal objectives, they will need to proceed in clearly defined stages. (author)

  19. Regulatory review of NPP Krsko Periodic Safety Review

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  1. Displacement Based Seismic Design Criteria

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. Structural concepts and details for seismic design

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Aziz, T.S.; Murakami, H.

    1991-01-01

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

  5. 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%

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  7. 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

  8. 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.)

  9. Design experience on seismically isolated buildings

    International Nuclear Information System (INIS)

    Giuliani, G.C.

    1991-01-01

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

  10. 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.

  11. Inelastic seismic behavior of post-installed anchors for nuclear safety related structures: Generation of experimental database

    Energy Technology Data Exchange (ETDEWEB)

    Mahadik, Vinay, E-mail: vinay.mahadik@iwb.uni-stuttgart.de; Sharma, Akanshu; Hofmann, Jan

    2016-02-15

    Highlights: • Experiments for evaluating seismic behavior of anchors were performed. • Two undercut anchor products in use in nuclear facilities were considered. • Monotonic tension, shear and cycling tension tests at different crack widths. • Crack cycling tests at constant, in-phase and out-of phase tension loads. • Characteristics for the two anchors as a function of crack width were identified. - Abstract: Post installed (PI) anchors are often employed for connections between concrete structure and components or systems in nuclear power plants (NPP) and related facilities. Standardized practices for nuclear related structures demand stringent criteria, which an anchor has to satisfy in order to qualify for use in NPP related structures. In NPP and related facilities, the structure–component interaction in the event of an earthquake depends on the inelastic behavior of the concrete structure, the component system and also the anchorage system that connects them. For analysis, anchorages are usually assumed to be rigid. Under seismic actions, however, it is known that anchors may undergo significant plastic displacement and strength degradation. Analysis of structure–component interaction under seismic loads calls for numerical models simulating inelastic behavior of anchorage systems. A testing program covering different seismic loading scenarios in a reasonably conservative manner is required to establish a basis for generating numerical models of anchorage systems. Currently there is a general lack of modeling techniques to consider the inelastic behavior of anchorages in structure–component interaction under seismic loads. In this work, in view of establishing a basis for development of numerical models simulating the inelastic behavior of anchors, seismic tests on two different undercut anchors qualified for their use in NPP related structures were carried out. The test program was primarily based on the DIBt-KKW-Leitfaden (2010) guidelines

  12. 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

  13. Safety design guides for seismic requirements for CANDU 9

    International Nuclear Information System (INIS)

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

    1996-03-01

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

  14. PROBABILISTIC SEISMIC ASSESSMENT OF BASE-ISOLATED NPPS SUBJECTED TO STRONG GROUND MOTIONS OF TOHOKU EARTHQUAKE

    Directory of Open Access Journals (Sweden)

    AHMER ALI

    2014-10-01

    Full Text Available The probabilistic seismic performance of a standard Korean nuclear power plant (NPP with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.

  15. Probabilistic seismic assessment of base-isolated NPPs subjected to strong ground motions of Tohoku earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Ahmer; Hayah, Nadin Abu; Kim, Doo Kie [Dept. of Civil and Environmental Engineering, Kunsan National University, Kunsan (Korea, Republic of); Cho, Sung Gook [R and D Center, JACE KOREA Company, Gyeonggido (Korea, Republic of)

    2014-10-15

    The probabilistic seismic performance of a standard Korean nuclear power plant (NPP) with an idealized isolation is investigated in the present work. A probabilistic seismic hazard analysis (PSHA) of the Wolsong site on the Korean peninsula is performed by considering peak ground acceleration (PGA) as an earthquake intensity measure. A procedure is reported on the categorization and selection of two sets of ground motions of the Tohoku earthquake, i.e. long-period and common as Set A and Set B respectively, for the nonlinear time history response analysis of the base-isolated NPP. Limit state values as multiples of the displacement responses of the NPP base isolation are considered for the fragility estimation. The seismic risk of the NPP is further assessed by incorporation of the rate of frequency exceedance and conditional failure probability curves. Furthermore, this framework attempts to show the unacceptable performance of the isolated NPP in terms of the probabilistic distribution and annual probability of limit states. The comparative results for long and common ground motions are discussed to contribute to the future safety of nuclear facilities against drastic events like Tohoku.

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

    International Nuclear Information System (INIS)

    Ign Djoko Irianto

    2004-01-01

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

  17. Analysis of design floor response spectra and testing of the electrical systems

    International Nuclear Information System (INIS)

    Ambriashvili, Y.

    1996-01-01

    This report covers the following activities as foreseen according to the working plan of 'Atmoenergoproject': analysis of calculated floor response spectra used during the design of Kozloduy NPP and comparison with other spectra recommended for this NPP; analysis of floor response spectrum for the most important systems (reactor, main coolant loop, electrical systems); tests of main electrical systems and analysis of the results on seismic stability of those systems. Results of the response spectra analysis are given, some of the electrical systems are identified by the Kozloduy authorities to be analyzed in future according to the results of the test on seismicity

  18. CONCIDERATION OF FOUNDATION AND SEISMIC CONDITIONS OF AREA IN ANALYSIS OF SEISMIC RESISTANCE OF REACTOR COMPARTMENT

    Directory of Open Access Journals (Sweden)

    SEDIN V. L.

    2015-11-01

    Full Text Available Problem statement. Providing of safe exploitation of nuclear power plants, as well as a safety of staff and environment is a very important problem. A distinct feature of this problem is a necessity to provide not only a strength of structures, but also a safe functioning of all systems that control nuclear process. In particular, the influence of earthquake should be considered on constructions of buildings and structures of nuclear and thermal power plant, taking into account soil-structure interaction. According to IAEA’s SSD-9 recommendations, a risk of vibration of soil should be analyzed for each NPP connected with earthquakes soil that means researches, including general, detailed and microseismic zoning of the area works. One of the distinctive features of the considered problem is an evaluation of the seismicity of area and getting the response spectrum on the free surface. Purpose. Determination of seismic resistance of buildings of high category of safety with the example of the reactor compartment of Zaporoghskaya NPP including the soil structure interaction. Conclusion The seismicity assessment of the area and obtaining of response specters on free surface was made during research and analysis of seismic resistance of buildings of high category of safety including the effects of foundation and structures. The method of modeling of the equivalent dynamic characteristics of the base was considered during the research in seismic impacts.

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

    International Nuclear Information System (INIS)

    Hara, Fumio; Iwasaki, Akira

    1996-01-01

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

  20. 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.

  1. Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-yeop; Lim, Ho-Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study.

  2. Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

    International Nuclear Information System (INIS)

    Kim, Sung-yeop; Lim, Ho-Gon

    2016-01-01

    Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study

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

    International Nuclear Information System (INIS)

    Zeman, P.

    1999-01-01

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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2000-04-01

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

  6. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

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

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

    Science.gov (United States)

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

    1981-05-01

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

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  9. 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

  10. Methodology and results of the seismic probabilistic safety assessment of Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Vermaut, M.K.; Monette, P.; Campbell, R.D.

    1995-01-01

    A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Krsko plant. The methodology adopted is the seismic PSA (Probabilistic Safety Assessment). The Krsko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of the site hazard, calculation of plant structures response including soil-structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Also relay chatter analysis and soil stability studies were performed. The seismic PSA described here is limited to the analysis of CDF (level I PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Krsko NPP but are not further described in this paper. The results of the seismic PSA study indicate that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable to that of most US and Western Europe NPPs. (author)

  11. An engineering design of reactor with NPP spent fuels

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Principles of Defense-in-depth philosophy applied in NPP engineering management

    International Nuclear Information System (INIS)

    Wu Guangwei

    2011-01-01

    Based on the Defense-in-depth Concept in nuclear and radiation safety, Defense-in-depth Concept for design management of Nuclear Power Plant (NPP) is developed in this paper to analyze the feasibility and importance of the application of the basic principle: Defense-in-depth concept in NPP systems performed during the design control of NPP. This paper focuses on the NPP engineering management process, and according to the analysis of such process, 5 principles of Defense-in-depth Concept applied in NPP design management are raised: (1) preventing the non-conformities of design via effective design quality management system; (2) discovering and correcting non-conformities of design quality in time via design checkup and design review meeting; (3) carrying out timely analysis and treatment against design non-conformities which have been transferred to construction phase; (4) Assessing and judging the severe non-conformities in construction phase, putting forward treatment opinions and remedies accordingly so as to avoid the existence of such non-conformities in physical construction of NPP; (5) Paying 'return-visit' and performing 'post-assessment' for NPP design to assess the designed functions and safety of NPP comprehensively. (author)

  14. 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

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

    International Nuclear Information System (INIS)

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

    2007-05-01

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

  16. Seismic fragility of ventilation stack of nuclear power plant

    International Nuclear Information System (INIS)

    Nefedov, S.S.; Yugai, T.Z.; Kalinkin, I.V.; Vizir, P.L.

    2003-01-01

    Fragility study of safety related elements is necessary step in seismic PSA of nuclear power plant (NPP). In present work fragility was analyzed after the example of the ventilation stack of NPP. Ventilation stack, considered in present work, is a separately erected construction with height of 100 m made of cast-in-place reinforced concrete. In accordance with IAEA terminology fragility of element is defined as conditional probability of its failure at given level of seismic loading. Failure of a ventilation stack was considered as development of the plastic hinge in some section of a shaft. Seismic ground acceleration a, which corresponds to failure, could be defined as limit seismic acceleration of ventilation stack [a]. Limit seismic acceleration [a] was considered as random value. Sources of its variation are connected with stochastic nature of factors determining it (properties of construction materials, soils etc.), and also with uncertainties of existing analytical techniques. Random value [a] was assumed to be distributed lognormally. Median m[a] and logarithmically standard deviation β of this distribution were defined by 'scaling method' developed by R.P. Kennedy et al. Using this values fragility curves were plotted for different levels of confidence probability. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    Science.gov (United States)

    Mwafy, Aman

    2011-03-01

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

  19. Seismic resistance design of nuclear power plant building structures in Japan

    International Nuclear Information System (INIS)

    Kitano, Takehito

    1997-01-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

  20. Seismic resistance design of nuclear power plant building structures in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Takehito [Kansai Electric Power Co., Inc., Osaka (Japan)

    1997-03-01

    Japan is one of the countries where earthquakes occur most frequently in the world and has incurred a lot of disasters in the past. Therefore, the seismic resistance design of a nuclear power plant plays a very important role in Japan. This report describes the general method of seismic resistance design of a nuclear power plant giving examples of PWR and BWR type reactor buildings in Japan. Nuclear facilities are classified into three seismic classes and is designed according to the corresponding seismic class in Japan. Concerning reactor buildings, the short-term allowable stress design is applied for the S1 seismic load and it is confirmed that the structures have a safety margin against the S2 seismic load. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. 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)

  3. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2B. General material: codes, standards, criteria. Working material

    International Nuclear Information System (INIS)

    1999-01-01

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

  4. 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)

  5. Key issues in european reactor seismic design

    International Nuclear Information System (INIS)

    Cicognani, G.; Martelli, A.

    1984-01-01

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

  6. Sensitivity of seismic design parameters to input variables

    International Nuclear Information System (INIS)

    Wium, D.J.W.

    1987-01-01

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

  7. New approaches to aseismic design and testing in Japan

    International Nuclear Information System (INIS)

    Ohsaki, Y.

    1980-01-01

    Problems of seismic safety and stability of the NPP components are discussed. Main provisions of safety standards being in force in Japan since 1978 are considered. Main specifications of the large-scale vibration-testing machine constructed for testing the mockups of the main components of NPPs with PWP and BWR reactors are given. In accordance with the rules of aseismic component design each element of a reactor must preserve its integrity under any possible seismic effects. Depending on the degree of radioactive contamination of environment resulted from the earthquake, the nuclear reactor equipment is divided into four classes. All buildings, accessories, pipelines and other equipment are also devided into four classes. The rules involve the estimation of seismic stability of components by the statical method on the base of the seismic stability coefficient established by the building standards. The method of ''time historical analysis'', in which the seismic design waves are used as input signals, is applied for the dynamic analysis of buildings; and for the equipment, pipelines and other elements the method of ''spectral analysis'' is used. In the future it is suggested to carry out the fundamental investigations of the NPP equipment safety and reliability, supported by corresponding experimental data, both from the point of view of arizing stresses and serviceability of all components directly at the moment of the earthquake beginning [ru

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

    International Nuclear Information System (INIS)

    Pentti Varpasuo

    2005-01-01

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

  9. Meeting up-to-date safety requirements in the Russian NPP projects

    International Nuclear Information System (INIS)

    Tepkyan, G. O.; Yashkin, A. V.

    2014-01-01

    Safety features in Russian NPP designs are implemented by the combination of active and passive safety systems • Russian NPP designs are in compliance with up-to-date international and European safety requirements and refer to Generation III+ • Russian state-of-the-art designs have already implemented some design solutions, which take into account “post-Fukushima” requirements. Russian NPP design principles have been approved during the European discussions in spring 2012, including the IAEA extraordinary session addressed to Fukushima NPP accident

  10. 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

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 5C. Experience data. Working material

    International Nuclear Information System (INIS)

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  13. Nonlinear Time Domain Seismic Soil-Structure Interaction (SSI) Deep Soil Site Methodology Development

    International Nuclear Information System (INIS)

    Spears, Robert Edward; Coleman, Justin Leigh

    2015-01-01

    methodology will be known as, NonLinear Soil-Structure Interaction (NLSSI). In general NLSSI analysis should provide a more accurate representation of the seismic demands on nuclear facilities their systems and components. INL, in collaboration with a Nuclear Power Plant Vender (NPP-V), will develop a generic Nuclear Power Plant (NPP) structural design to be used in development of the methodology and for comparison with SASSI. This generic NPP design has been evaluated for the INL soil site because of the ease of access and quality of the site specific data. It is now being evaluated for a second site at Vogtle which is located approximately 15 miles East-Northeast of Waynesboro, Georgia and adjacent to Savanna River. The Vogtle site consists of many soil layers spanning down to a depth of 1058 feet. The reason that two soil sites are chosen is to demonstrate the methodology across multiple soil sites. The project will drive the models (soil and structure) using successively increasing acceleration time histories with amplitudes. The models will be run in time domain codes such as ABAQUS, LS-DYNA, and/or ESSI and compared with the same models run in SASSI. The project is focused on developing and documenting a method for performing time domain, non-linear seismic soil structure interaction (SSI) analysis. Development of this method will provide the Department of Energy (DOE) and industry with another tool to perform seismic SSI analysis.

  14. Seismic Observation in Deep Boreholes and Its Applications - Workshop Proceedings, Niigata Institute of Technology, Kashiwazaki, Japan

    International Nuclear Information System (INIS)

    2014-01-01

    The Kashiwazaki WS was held to develop the recommendations made at the Tsukuba WS entitled 'Seismic Input Motions Incorporating Recent Geological Studies' which was held in November 2004 in Tsukuba City in Japan (hereinafter, the 'Tsukuba WS'). At the Tsukuba WS, the state of the art in defining realistic seismic input for the design and re-evaluation of nuclear facilities as well as advances in seismic source characterization of fault zones using data from deep geological investigations and their possible contribution to improving seismic input definitions were reviewed. Further, the importance and necessity of cooperation between seismology and geology in order to decrease uncertainty in seismic input definition were emphasized. After the Tsukuba WS, the Niigata-ken Chuetsu-oki Earthquake (NCOE, M=6.8) occurred near the Kashiwazaki-Kariwa NPP site. In this earthquake, a focusing effect of ground motion was observed at the Kashiwazaki-Kariwa NPP site, which caused locally amplified ground motion. Units 1 and 2 showed significantly higher responses of more than 50 percent compared to Units 3 to 7. This was thought to be caused by the fact that seismic waves were focused on Units 1 and 2 due to the irregular structure under the site. A similar effect was observed at the Hamaoka site in the earthquake that occurred in Suruga Bay in 2009. Only Unit 5 showed a double or larger response to this earthquake, and similar phenomena were found only for events with hypocenter in the narrow direction from the site. This was also thought to be caused by the irregular geological structure under the site. In the 2011 off the Pacific Coast of Tohoku Earthquake, the peak ground accelerations (PGA) of the Dai-ichi NPP site were about twice as large as those of the Dai-ni NPP site, although the distance between these sites is only slightly more than 10 km. Further, large differences in the PGAs were found among each of the units of the Dai-ichi NPP site. For example, the PGA of Unit

  15. 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

  16. Seismic functional qualification of active mechanical and electrical components based on shaking table testing

    International Nuclear Information System (INIS)

    Jurukovski, D.

    1999-01-01

    The seismic testing for qualification of one sample of the NPP Kozloduy Control Panel type YKTC was carried out under Research Contract no: 8008/Rl, entitled: 'Seismic Functional Qualification of Active Mechanical and Electrical Components Based on Shaking Table Testing'. The tested specimen was selected by the Kozloduy NPP staff, Section 'TIA-2' (Technical Instrumentation and Automatics), however the seismic input parameters were selected by the NPP Kozloduy staff, Section HTS and SC (Hydro-Technical Systems and Engineering Structures). The applied methodology was developed by the Institute of Earthquake Engineering and Engineering Seismology staff. This report presents all relevant items related to the selected specimen seismic testing for seismic qualification such as: description of the tested specimen, mounting conditions on the shaking table, selection of seismic input parameters and creation of seismic excitations, description of the testing equipment, explanation of the applied methodology, 'on line' and 'off line' monitoring of the tested specimen, functioning capabilities, discussion of the results and their presentation and finally conclusions and recommendations. In this partial project report, two items are presented. The first item presents a review of the existing and used regulations for performing of the seismic and vibratory withstand testing of electro-mechanical equipment. The selection is made based on MEA, IEEE, IEC and former Soviet Union regulations. The second item presents the abstracts of all the tests performed at the Institute of Earthquake Engineering and Engineering Seismology in Skopje. The selected regulations, the experience of the Institute that has been gathered for the last seventeen years and some theoretical and experimental research will be the basis for further investigations for development of a synthesised methodology for seismic qualification of differently categorized equipment for nuclear power plants

  17. 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)

  18. 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.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. Transition cycle fuel management problems of NPP Krsko

    International Nuclear Information System (INIS)

    Petrovic, B.; Pevec, D.; Smuc, T.; Urli, N.

    1989-01-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

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

    International Nuclear Information System (INIS)

    Barariu, Gheorghe

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Benzoni, G.; Seible, F.

    1998-01-01

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

  4. Treatment of NPP wastes using vitrification

    International Nuclear Information System (INIS)

    Sobolev, I.A.; Lifanov, F.A.; Stefanovsky, S.V.; Kobelev, A.P.; Savkin, A.E.; Kornev, V.I.

    1998-01-01

    Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10 -5 -10 -6 g/(cm 2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

  5. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

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

  6. 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

  7. Experiences in seismic upgrading of equipment and structures in Kozloduy nuclear power plant (440 WWER-PWR)

    International Nuclear Information System (INIS)

    Ordonez Villalobos, A.

    1993-01-01

    Within the framework of the 'Emergency programme for Nuclear Safety of Kozloduy NPP' it has been concluded that the increase in seismic safety of a NPP can be achieved by upgrading the key equipment in a cost effective way. Essential and vulnerable equipment has to be identified. Seismic capacity should be evaluated base don realistic state of the art criteria. Seismic review teams ef experienced engineers should conduct planned walk-downs in order to propose effective upgrading solutions. Team work of plan engineers and construction engineers would enhance the effectiveness of the solutions. It is recommended that all the participants be motivated and have a clear understanding of the objectives of the upgrading

  8. Seismic Design Guidelines For Port Structures

    DEFF Research Database (Denmark)

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

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

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

    International Nuclear Information System (INIS)

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

    1989-02-01

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

  10. Research on the NPP human factors engineering operating experience review

    International Nuclear Information System (INIS)

    Ren Xiangchen; Miao Hongxing; Ning Zhonghe

    2006-01-01

    This paper addresses the importance of the human factors engineering (HFE) for the design of nuclear power plant (NPP), especially for the design of human-machine interface in the NPP. It also summarizes the scope and content of the NPP HFE. The function, scope, content and process of the NPP human factors engineering operating experience review (OER) are mainly focused on, and significantly discussed. Finally, it briefly introduces the situation of the studies on the OER in China. (authors)

  11. Ageing management database development for PWR NPP steam generator

    International Nuclear Information System (INIS)

    Liu Hongyun; Xu Liangjun; Xiong Changhuai; Wang Xianyuan

    2005-01-01

    Steam generator (SG) is one of the key safe important equipment of NPP, which is covered by NPP aging management program. Steam Generator Aging Management Dabatase (SGAMDB) is developed to provide necessary information for SG aging management. RINPO is developing SGAMDB for domestic NPP. This system contains information and data about SG design, manufacture, operation and maintenance. The information include NPP fundamental data, SG design data, SG aging mechanism, SG operation data, SG ISI data, SG maintenance data and SG evaluation interface. The system runs at the intranet of Qinshan-1 NPP with B/S mode. It can provide information inquire and fundamental analysis for NPP SG aging team and SG aging researcher's. In addition, it provides necessary information and data for SG aging analysis and evaluation, such as all pressure test process and flaws of tubes, and collects the analysis results. (authors)

  12. 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

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

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1993-01-01

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

  15. Seismic qualification of existing nuclear installations in India - a proposal

    International Nuclear Information System (INIS)

    Basu, P.C.

    2001-01-01

    In India, the work toward seismic qualification of existing nuclear facilities has been started. Preliminary work is being undertaken with respect to identifying the facilities which would be taken up for seismic qualification, approach and methodology for re-evaluation for seismic safety, acceptance criteria, etc. Work has also been started for framing up the criteria and methodology of the seismic qualification of these facilities. Present paper contains the proposal in this respect. This proposal is on similar lines of the present practice of seismic qualification of NPP, as summarized in the Appendix, but has been modified to suit the special requirements of Indian nuclear installations. (author)

  16. Operation Aspect of the Main Control Room of NPP

    International Nuclear Information System (INIS)

    Sahala M Lumbanraja

    2009-01-01

    The main control room of Nuclear Power Plant (NPP) is operational centre to control all of the operation activity of NPP. NPP must be operated carefully and safely. Many aspect that contributed to operation of NPP, such as man power whose operated, technology type used, ergonomic of main control room, operational management, etc. The disturbances of communication in control room must be anticipated so the high availability of NPP can be achieved. The ergonomic of the NPP control room that will be used in Indonesia must be designed suitable to anthropometric of Indonesia society. (author)

  17. Seismic performance assessment of base-isolated safety-related nuclear structures

    Science.gov (United States)

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

    2010-01-01

    Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

  18. Contribution of the deterministic approach to the characterization of seismic input

    International Nuclear Information System (INIS)

    Panza, G.F.; Romanelli, F.; Vaccari, F.; Decanini, L.; Mollaioli, F.

    1999-10-01

    due to the intermediate -depth earthquakes, whereas the deterministic results seem representative of the most conservative scenario. The peak values of D, V and DGA, and pertinent periods T(D) and T(V), at the sites where nuclear power plants are located are summarized in a table. Given the number of Nuclear Power Plants (NPP) located in the studied region, the obtained results should be refined by successive more detailed investigations aimed at the retrofitting of the existing NPP. This may be a necessary action in order to reduce the environmental hazard associated with such NPP. The realistic modeling of the ground motion can be used to generate a very important base of knowledge for the preparation of ground shaking scenarios and it is at the base of the UNESCO IGCP Project 414 'Realistic Modeling of Seismic Input for Megacities and Large Urban Areas'. The knowledge can be very fruitfully used by civil engineers in the design of new seismo-resistant constructions and in the reinforcement of the existing built environment, and, therefore, supply a particularly powerful tool for the prevention aspects of Civil Defence. The results of the deterministic approach can be immediately used for the estimation of the destructive potential from (a) the direct analysis of the calculated signals, (b) their integration in time or in frequency domain, and (c) the calculation of the corresponding structural (elastic and anelastic) response. Examples of application of this analysis to urban areas in eastern Sicily, where more than 700.000 people are living, show that the use of advanced seismological methods for the definition of ground shaking scenarios can play a key role in the reduction of possible losses. (author)

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

    International Nuclear Information System (INIS)

    Baliza, Ana Rosa

    2013-01-01

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

  20. Organization of monitoring of agricultural products in NPP region

    International Nuclear Information System (INIS)

    Panteleev, L.I.; Spirin, E.V.; Sanzharova, N.I.

    1990-01-01

    Problem of organizing chemical and radiation monitoring of agricultural products in NPP region is considered. Attention is paid to monitoring during NPP siting and designing, to monitoring of radioactive contamination of agricultural products under normal NPP operation, emergency situations and decommissioning

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

    International Nuclear Information System (INIS)

    1996-01-01

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

  2. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  3. 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)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  5. Calculational experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs. Final report

    International Nuclear Information System (INIS)

    1999-01-01

    The results of testing of equipment at Bohunice NPP and pipeline systems at Unit 3 of Kozloduy NPP (WWER-440 type reactors) are presented in this Final Report. These results side by side with experimental values of natural frequencies and decrements also include experimental data about vibration modes of tested equipment and pipelines. For the first time the results of new calculational-experimental examination of equipment seismic resistance at Unit 2 of Armenian NPP are presented. At Kozloduy NPP direction's request the planed additional tests of some selected items were put off on 1997. Instead of postponed tests we carried out detailed analysis of our past inspections of numerous equipment seismic resistance at the Unit 5 of Kozloduy NPP. Experimental data with results of additional analysis are presented

  6. Seismic hazard maps for earthquake-resistant construction designs

    International Nuclear Information System (INIS)

    Ohkawa, Izuru

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Kunar, R.R.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Guerpinar, A.; Zola, M.

    2001-01-01

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

  9. 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.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Reddy, G.R.

    2011-01-01

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

  12. User's manual of SECOM2: a computer code for seismic system reliability analysis

    International Nuclear Information System (INIS)

    Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

    2002-03-01

    This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

  13. Civil Works Seismic Designs

    International Nuclear Information System (INIS)

    1985-12-01

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

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

    International Nuclear Information System (INIS)

    1980-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

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

  16. Application of mathematical model for high viscous damper to dynamic analysis of NPP pipings

    International Nuclear Information System (INIS)

    Kostarev, V.V.; Bercovsky, A.M.; Kireev, O.B.; Vasiliev, P.S.

    1993-01-01

    The problems of dynamic analysis of Nuclear Power Plants (NPP) piping systems are considered in the paper. The special calculation program for PC has been developed that enables to estimate the seismic margin for any piping system with different antiseismic devices having nonlinear characteristics. The calculated comparison has been done for two antiseismic supports that are widely used now, namely: a High Viscous Damper (HVD) and a Seismic Stop Support (SSS) with the application, as an example, to the well known pipeline BM3 (USNRC). (author)

  17. Application of mathematical model for high viscous damper to dynamic analysis of NPP pipings

    Energy Technology Data Exchange (ETDEWEB)

    Kostarev, V V; Bercovsky, A M; Kireev, O B; Vasiliev, P S [CKTI VIBROSEISM (CVS), St. Petersburg (Russian Federation)

    1993-07-01

    The problems of dynamic analysis of Nuclear Power Plants (NPP) piping systems are considered in the paper. The special calculation program for PC has been developed that enables to estimate the seismic margin for any piping system with different antiseismic devices having nonlinear characteristics. The calculated comparison has been done for two antiseismic supports that are widely used now, namely: a High Viscous Damper (HVD) and a Seismic Stop Support (SSS) with the application, as an example, to the well known pipeline BM3 (USNRC). (author)

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

    International Nuclear Information System (INIS)

    Cha, Kyung H.

    2011-01-01

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

  19. 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)

  20. Seismic-design questions typify nuclear obstacles

    International Nuclear Information System (INIS)

    Strauss, S.D.

    1979-01-01

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

  1. Akkuyu NPP – the first Turkish NPP. The new history of the project

    International Nuclear Information System (INIS)

    Tzocheva, V.

    2012-01-01

    An overview is given to the Turkish energy sector and nuclear power plans. The project for the construction of the first NPP in Turkey is presented. The general parameters of the Project are: CAPEX: $ 20 bln; Project design: NPP-2006; (VVER- 1200); Number of units: 4; Total capacity: 4 800 MW; Construction period: 2014 – 2023; PPA period; 15 years, fixed price terms. An account of the activities during 2011, the Worley Parsons participation are presented and a tentative project schedule is given

  2. A review on design and study of floating plant mooring systems

    International Nuclear Information System (INIS)

    Nakamura, Hideharu; Kashima, Ryoichi; Hagiwara, Yutaka; Matsuura, Shinichi; Shiojiri, Hiroo

    1985-01-01

    There exists a widespread anticipation among nuclear power engineers that a floating nuclear power plant (NPP) may have a couple of advantages over other types; such as seismic isolation, standardizations of design and manufacturing etc. However, it also seems that a number of problems still remain to be clarified to adopt the floating NPP in Japan. One of them is magnitude of external forces acting on floating structure, which depend on mooring system. The purpose of the present report is to survey various kinds of mooring systems, structural analyses and design criteria, and examples of the same scale floating structures taking into consideration of the floating NPP which displaces 300,000 Ton, and furthermore, to discuss the future problems in regard to the mooring system still to be investigated before the floating plants can be constructed. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  4. Seismic analyses of structures. 1st draft

    International Nuclear Information System (INIS)

    David, M.

    1995-01-01

    The dynamic analysis presented in this paper refers to the seismic analysis of the main building of Paks NPP. The aim of the analysis was to determine the floor response spectra as response to seismic input. This analysis was performed by the 3-dimensional calculation model and the floor response spectra were determined for a number levels from the floor response time histories and no other adjustments were applied. The following results of seismic analysis are presented: 3-dimensional finite element model; basic assumptions of dynamic analyses; table of frequencies and included factors; modal masses for all modes; floor response spectra in all the selected nodes with figures of indicated nodes and important nodes of free vibration

  5. Seismic analyses of structures. 1st draft

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

    The dynamic analysis presented in this paper refers to the seismic analysis of the main building of Paks NPP. The aim of the analysis was to determine the floor response spectra as responseto seismic input. This analysis was performed by the 3-dimensional calculation model and the floor response spectra were determined for a number levels from the floor response time histories and no other adjustments were applied. The following results of seismic analysis are presented: 3-dimensional finite element model; basic assumptions of dynamic analyses; table of frequencies and included factors; modal masses for all modes; floor response spectra in all the selected nodes with figures of indicated nodes and important nodes of free vibration.

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

    International Nuclear Information System (INIS)

    Rodriguez Aleman, Carlos; Mitjans Sanchez, Guillermo

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Seismic design criteria for special isotope separation plant structures

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.

    1998-01-01

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

  10. 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

  11. Research contract: Seismic stability of nuclear power plants in Eastern Europe. Final report. From 15 April 1998 to 15 April 1999

    International Nuclear Information System (INIS)

    Ambriashvili, Y.

    1999-01-01

    General scientific scope of the presented program is assessment of stability and functionality of the nuclear power plants with RBMK type reactors in relation to External Evens including following: seismic capacity of structures, equipment and distribution systems; capacity of structures for impact type loading; capacity of structures for blast type loading. For the analyses only structures, equipment and distribution systems which are responsible for safety shutdown path are used. Since 1980 Atomenergoproject has been participated in development and carrying out the Research Program related to investigation of seismic stability of RBMK NPPs. This investigation was done for Smolensk and Kursk NPPs. It is known that the design basis of seismic analyses is investigation of dynamic characteristic (main frequencies and main modes) of structures, equipment and distribution systems. Therefore the assessment of capacity of structures and systems can be based on the results of seismic stability investigations. In the present final report some main results of dynamic analyses reactor building, electrical buildings, storage building, pipe lines, separators, electrical equipment, etc. are described. On the basis of the data site seismological investigation, calculations and different type of testing structures and equipment it was accepted, that the RBMK NPP is the Seismic Stability NPP up to 6 intensity by MSK64 scale. The results of calculation and testing investigations can be use for the dynamic analyses on the external events blast and impact loading. Using the results the probability analysis should be done

  12. Integrated ageing management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos; Sabransky, Mario

    2013-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  13. Integrated Ageing Management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, J.M.; Marchena, M.H.; Zorrilla, J.R.; Sabransky, M.

    2012-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor twice as big as Atucha I finishing a delayed construction . With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  14. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos, E-mail: ranalli@cnea.gov.ar [Gerencia Coordinacion Proyectos CNEA-NASA, Comision Nacional de Energia Atomica, Buenos Aires (Argentina); Sabransky, Mario, E-mail: msabransky@na-sa.com.ar [Departamento Gestion de Envejecimiento, Central Nuclear Atucha I-II Nucleoelectrica Argentina S.A., Provincia de Buenos Aires (Argentina)

    2013-07-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Design Guidelines of a Spring-Damper System for Emergency Diesel Generator Sets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Kyu; Choun, Young Sun; Seo, Jeong Moon

    2007-05-15

    This guidelines described about the procedure of isolation system design for Emergency Diesel Generator (EDG) of Nuclear Power Plant (NPP). First of all, a vibration concept including the ground vibration was described and vibration control system and seismic isolation system were considered. The behavior characteristics and design consideration of coil spring-viscose damper system were summarized. The material properties of foundation of EDG system and the ground were considered. A design load and seismic load for isolation system design were described and an analysis method was explained. Finally, a design example for an EDG in Yonggwang Nuclear Unit 5 and 6 was attached of Appendix. First of all, this design guideline can apply to design of a vibration and seismic isolation system for EDG system and the design example present a design procedure practically. Moreover, this design guideline can be used for isolation design of other rotational machines and other isolation system except spring-damper system.

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-05-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. Transposition of the SQUG methodology to the Belgian NPP(nuclear power station)

    International Nuclear Information System (INIS)

    Detroux, P.; Aelbrecht, D.; Naisse, J.C.; Greer, J.L.

    1991-01-01

    The units of a Belgian Nuclear PowerStation had to be seismically reassessed after ten years of operation because the seismic requirements were upgraded from 0.1g to 0.17g free field ground acceleration. Seismic requalification of the active equipment was a critical problem as the current classical methods were too conservative and their application would have lead to unacceptable replacement or reinforcement of a lot of equipment. The approach based on the use of past experience of seismic behavior of non nuclear equipment was chosen; this methodology was developed by the Seismic Qualification Utility Group (SQUG); a group of U.S. utilities and had to be transposed to the Belgian N.P.P. This transposition is described in this paper. It affects different aspects of the methodology. First, the impact of specific requests of the Safety Authorities on the elaboration of the Safe Shutdown Equipment List (SSEL) shall be examined. Then it is explained why the tedious work of specific relay screening was avoided by taking advantage of initial design features for both Instrumentation and Control (I and C) and Electrical power distribution system; the impact on the Electrical SSEL is also described. Afterwards, it is presented how it was possible to conduct a specific existing seismic qualification at 0.1 g free field ground acceleration. Finally, the resolution of specific important problems that arose from the application of the SQUG methodology, is presented such as the definition of the grade level and the conservatism of the classical Amplified Floor Spectra (criterion 1), the calculation of the nozzle loads on mechanical equipment connected to long unbraced piping and the transfer of these loads to the anchorage. (author)

  1. Leningrad NPP and energetics of north-western Russia

    International Nuclear Information System (INIS)

    Belov, I.

    2000-01-01

    Problems of Leningrad NPP operating units modernization, their design service life finishing by year of 2010, are discussed. To assure safe operation of unit 1 investments in the amount of 30 mln. dol are necessary. Estimations suggest economic efficiency of the measures, permitting saving of 300 mln. dol worth of gas. Unfortunately, without a rise in tariff for electric power produced by NPP it seems impossible. It is recommended that substantiated tariffs are set for electric power produced by NPP starting from January, 2000. The measure is indispensable for raising investment funds intended for operating NPP modernization [ru

  2. Seismic design practice for Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Calculating seismic of slabs ITA NNP Garona

    International Nuclear Information System (INIS)

    Ezeberry, J. I.; Guerrero, A.; Gamarra, J.; Beltran, F.

    2014-01-01

    This article describes the methodology that Idom has employed to perform the seismic evaluation of slabs within the ITA project of the NPP Santa Maria de Garona. Seismic calculations that have been conducted include consideration of the effects of the interaction of soil structure as well as the possible take-off containers with respect to slab during the earthquake. Therefore, the main contribution of the work is the study of the coupling of rolling containers with the flexibility of the whole ground-slab For calculations has been used ABAQUS/Explicit program, allowing to solve effectively the nonlinearities listed above using explicit integration algorithms over time. The results of the calculations reflect the importance of jointly analyse the seismic responses of slab and containers. (Author)

  5. Seismic design considerations of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2001-10-01

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

  6. Technical innovations at NPP Dukovany - for safe and efficient operation

    International Nuclear Information System (INIS)

    Sabata, M.; Vasa, I.

    2000-01-01

    Inherent features of the NPP Dukovany design provide a significant confidence in its nuclear safety assurance; among these features should be emphasised the reactor core stability and its control and protection system capability to hold the reactor in safe state following scram or accident conditions. Nevertheless, NPP Dukovany was designed in the early seventies, and current requirements for nuclear safety assurance are more strict and/or specific as a result of the technical development and lessons learned from nuclear accidents. The paper compares the safety design base established at the time of NPP Dukovany project implementation and the current reference design base. The paper also presents procedures applied to implement technical and operational measures which are introduced to fulfil the current basic safety criteria. The scope of such measures applied at NPP Dukovany is compared with that of EU countries introduced for the same reason - to meet the updated safety related requirements. Examples of some innovations already implemented or under implementation give an idea how NPP Dukovany proceeds in reaching the goal of harmonising its safety with the requirements to be met before the Czech Republic becomes a member country of the European Union. (author)

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

    Science.gov (United States)

    2013-11-01

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

  8. 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

  9. Review report on the dynamical study of the main building of the Paks NPP

    International Nuclear Information System (INIS)

    Gatti, F.

    1995-01-01

    The present report deals with the review of the report 'Dynamical Study of the main building of the Paks NPP', issued by Paks NPP (Hungary) on April, 1993, within the frame of the IAEA benchmark study for the seismic analysis and testing of an existing Nuclear Power Plant (M), and on behalf of ENEL DSR/VDN Rome, in the aims of the nuclear activities of ENEL DSR/VDN (Rome). After a foreword to define the aims of the job (Chapter 1) and the identification of the scope of the work (Chapter 2), a short list of references is given (Chapter 3). In Chapter 4, the criteria followed in the review activity are listed; in Chapter 5, the contents of the Paks NPP report are summarized. In Chapter 6 the results of the review are given, while the main conclusions of the review activities are summarized in the Chapter 7. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Science.gov (United States)

    Veeraian, Parthasarathi; Gandhi, Uma; Mangalanathan, Umapathy

    2018-04-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. 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

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

    Directory of Open Access Journals (Sweden)

    Ahmer Ali

    2017-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-15

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

  16. 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.)

  17. 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.)

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

    International Nuclear Information System (INIS)

    Terasvirta, R.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Gong Zhenbang; Zhang Renyan

    2014-01-01

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

  1. Recent development of seismic evaluation for Swedish NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Bennemo, L [Vattenfall Energisystem, Stockholm (Sweden)

    1997-03-01

    In Scandinavia seismic activity is generally low. Only a few incidents have been registered in historic time, which might have damaged an industrial plant of today. There has been no earthquakes in Sweden strong enough to affect a NPP during our nuclear era (and not for very long time before either). So the risk for an nuclear accident i Sweden, caused by an earthquake, may thus be considered to be low. The basis and the methodology used in the design of Forsmark 3 and Oskarshamn 3 with respect to seismic safety is not in all parts suited to be employed for the older reactors. The methods implies a number of simplifications which may be a practical approach in connection with a new design but which might cause too conservative judgements of existing designs. The development of methods is therefore a vital part in the analysis. The Swedish nuclear Power Inspectorate (SKI), Vattenfall AB, Sydkraft AB and Oskarshams Kraftgrupp AB (OKG) have performed such a development of methods in a joint research program: `Project Seismic Safety`. The aim of the project was to develop methods for calculating the ground response to be used in the safety analysis of nuclear power plants in Sweden, as well as to demonstrate its application to the power plants at Ringhals and Barseback. The study also included a survey of geological and seismological conditions in the regions around the power plants studied. Since the large scale geological and seismological conditions around the individual nuclear plant sites are not very different as regards their expected effects on the seismic ground motion, the results obtained for the `typical hard rock site` can be taken as a basis for the characterization of the ground motions at the individual sites, after appropriate transformations to account for specific load conditions, seismological as well as geological. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. Safety enhancement concept for NPP of new generation with VVER reactors

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Kukhtevich, I.; Semashko, S.; Svetlov, S.; Solodovnikov, A.

    2004-01-01

    With the present day conditions, in order to successfully promote new NPP designs in the electric power markets, it is necessary to ensure enhanced technical/economic performances provided that international safety requirements are properly adhered to. When compared with high-powered nuclear power plants, NPP VVER-640 design (medium powered) possesses a number of advantages for the regions with undeveloped energy systems. Reduced specific energy intensity of the core adopted in this type of reactor allows to ensure the emergency cooldown of the reactor plant by passive means and to minimize the 'human factor' risk and external effects and provide sound substantiations as to how to retain corium inside RPV in case of severe accidents. At the same time, high-powered NPPs seem to be promising for regions with developed energy systems. Among such designs, NPP VVER-1000 and VVER-1500 designs are the most desirable. Configuration of new generation NPP with VVER-1500 is to be selected based on the gained experience in designing NPPs of previous generations considering the latest safety requirements and situation in the domestic and global energy markets for the time being and in the short run. Recent IAEA publications and latest EUR requirements insist that the following key safety indices should be established for new NPP designs: - aggregated frequency of core melting is 10 -6 (1/year); - frequency of maximum accident release is 10 -7 (1/year). To meet the aforementioned criteria, it is necessary to implement some safety assurance principles recommended by IAEA (in-depth defence, single failure, redundancy, diversity, etc.), application of deterministic and probabilistic methods for selection of safety assurance activities and means and use of reasonable combination of active and passive systems. Application of VVER-640 concept to high-powered NPPs seems to be a formidable task due to a number of reasons, namely, it is quite difficult to carry out cooldown process

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. Probabilistic Seismic Hazard Assessment Method for Nonlinear Soil Sites based on the Hazard Spectrum of Bedrock Sites

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Seo, Jeong Moon; Choi, In Kil

    2011-01-01

    For the probabilistic safety assessment of the nuclear power plants (NPP) under seismic events, the rational probabilistic seismic hazard estimation should be performed. Generally, the probabilistic seismic hazard of NPP site is represented by the uniform hazard spectrum (UHS) for the specific annual frequency. In most case, since that the attenuation equations were defined for the bedrock sites, the standard attenuation laws cannot be applied to the general soft soil sites. Hence, for the probabilistic estimation of the seismic hazard of soft soil sites, a methodology of probabilistic seismic hazard analysis (PSHA) coupled with nonlinear dynamic analyses of the soil column are required. Two methods are commonly used for the site response analysis considering the nonlinearity of sites. The one is the deterministic method and another is the probabilistic method. In the analysis of site response, there exist many uncertainty factors such as the variation of the magnitude and frequency contents of input ground motion, and material properties of soil deposits. Hence, nowadays, it is recommended that the adoption of the probabilistic method for the PSHA of soft soil deposits considering such uncertainty factors. In this study, we estimated the amplification factor of the surface of the soft soil deposits with considering the uncertainties of the input ground motions and the soil material properties. Then, we proposed the probabilistic methodology to evaluate the UHS of the soft soil site by multiplying the amplification factor to that of the bedrock site. The proposed method was applied to four typical target sites of KNGR and APR1400 NPP site categories

  8. Rigid-plastic seismic design of reinforced concrete structures

    DEFF Research Database (Denmark)

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

    2007-01-01

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

  9. 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

  10. Development of Probabilistic Performance Evaluation Procedure for Umbilical Lines of Seismically Isolated NPPs

    International Nuclear Information System (INIS)

    Hahm, Daegi; Park, Junhee; Choi, Inkil

    2013-01-01

    In this study, we proposed a procedure to perform the probabilistic performance evaluation of interface piping system for seismically isolated NPPs, and carried out the preliminary performance evaluation of the target example umbilical line. For EDB level earthquakes, the target performance goal cannot be fulfilled, but we also find out that the result can be changed with respect to the variation of the assumed values, i. e., the distribution of response, and the limit state of piping system. Recently, to design the nuclear power plants (NPPs) more efficiently and safely against the strong seismic load, many researchers focus on the seismic isolation system. For the adoption of seismic isolation system to the NPPs, the seismic performance of isolation devices, structures, and components should be guaranteed firstly. Hence, some researches were performed to determine the seismic performance of such items. For the interface piping system between isolated structure and non-isolated structure, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. Nowadays, in NUREG report, the probabilistic performance criteria for isolated NPP structures and components are proposed. Hence, in this study, we developed the probabilistic performance evaluation method and procedure for interface piping system, and applied the method to an example pipe. The detailed procedure and main results are summarized in next section. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system

  11. 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

  12. 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 ...

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

    International Nuclear Information System (INIS)

    Ito, Tomohiro; Michiue, Masashi; Fujita, Katsuhisa

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Sollogoub, P.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Alderson, M.A.H.G.

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1985-04-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-08-01

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

  20. German seismic regulations

    International Nuclear Information System (INIS)

    Danisch, Ruediger

    2002-01-01

    Rules and regulations for seismic design in Germany cover the following: seismic design of conventional buildings; and seismic design of nuclear facilities. Safety criteria for NPPs, accident guidelines, and guidelines for PWRs as well as safety standards are cited. Safety standards concerned with NPPs seismic design include basic principles, soil analysis, design of building structures, design of mechanical and electrical components, seismic instrumentation, and measures to be undertaken after the earthquake

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

    Science.gov (United States)

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

    2018-01-01

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

  2. Level 1 shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

    International Nuclear Information System (INIS)

    Halada, P.; Cillik, I.; Stojka, T.; Kuzma, M.; Prochaska, J.; Vrtik, L.

    2004-01-01

    The paper presents general approach, used methods and form of documentation of the results that have been applied within the shutdown and low power PSA (SPSA) study for Mochovce NPP, Unit 1, Slovakia. The SPSA project was realized by VUJE Trnava Inc., Slovakia in 2001-2002 years. The Level 1 SPSA study for Mochovce NPP Unit 1 covers internal events as well as internal (fires, floods and heavy load drop) and external (aircraft crash, extreme meteorological conditions, seismic event and influence of surrounding industry) hazards. Mochovce NPP consists of two operating units equipped with VVER 440/V213 reactors safety upgraded before construction finishing and operation start. 87 safety measures based on VVER 440 operational experience and international mission insights were implemented to enhance its operational and nuclear safety. The SPSA relates to full power PSA (FPSA) as a continuation of the effort to create a harmonized level 1 PSA model for all operational modes of the plant with the goal to use it for further purposes as follows: Real Time Risk Monitor, Maintenance Optimization, Technical Specifications Optimization, Living PSA. (author)

  3. Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

    International Nuclear Information System (INIS)

    Guilin, L.; Defu, L.; Huajun, L.; Fengqing, W.; Tao, Z.

    2012-01-01

    With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

  4. Integrated seismic design of structure and control systems

    CERN Document Server

    Castaldo, Paolo

    2014-01-01

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

  5. Project No. 7 - Decommissioning unit at Ignalina NPP. (Engineering and project management)

    International Nuclear Information System (INIS)

    2000-01-01

    In order to manage decommissioning process at Ignalina NPP, a new unit should be established in the framework of the Ignalina NPP organizational structure. The on-site Engineering Project Management Unit (PMU) will adopt an integrated approach to project management , engineering design, planning, procurement, safety and licensing activities at Ignalina NPP site. The PMU will assist the Ignalina NPP management in the development of an integrated decommissioning and waste management strategy

  6. Innovative design of viscoelastic dampers for seismic mitigation

    International Nuclear Information System (INIS)

    Tsai, C.S.

    1993-01-01

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

  7. Seismic considerations in the design of atomic power plants

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  8. Dissemination of Knowledge about NPP Instrumentation and Control Systems

    International Nuclear Information System (INIS)

    Yastrebenetsky, M.

    2016-01-01

    Full text: Instrumentation and control (I&C) systems are the most variable part in the nuclear power plants (NPP) comparatively with any other NPP systems. This statement is connected with the wide use of computers, rapid changes in information technologies, with the appearance of new computer complex electronic components, e.g., field programmable gate arrays (FPGA) and with appropriate point of their insertion into NPP I&C life cycle. The changes in NPP I&C systems require the dissemination of the knowledge about these systems. Lessons after Fukushima accident increase necessity of these actions. The elaboration and following dissemination of this knowledge took place in different directions: • Writing and issue of three new books about NPP I&C systems for specialists and for students which were issued in Ukrainian and USA public houses (the last book was issued in 2014); • Organization of five international scientific technical conferences, devoted to NPP I&C safety problems; • Elaboration of national (Ukrainian) standards and regulations pertaining to safety important NPP I&C systems (the last standard was issued in 2015) and participation in elaboration of international standards; • Lecturing for university students, NPP specialists and I&C designers. These actions in all directions are added to IAEA activity in the area NPP I&C systems (e.g., IAEA NP-T-3.12 “Core Knowledge on I&C systems in NPP”). (author

  9. Radiation monitoring in the NPP environment, control of radioactivity in NPP-environment system

    International Nuclear Information System (INIS)

    Egorov, Yu.A.

    1987-01-01

    Problems of radiation monitoring and control of the NPP-environment system (NPPES) are considered. Radiation control system at the NPP and in the environment provides for the control of the NPP, considered as the source of radioactive releases in the environment and for the environmental radiation climate control. It is shown, that the radiation control of the NPP-environment system must be based on the ecological normalization principles of the NPP environmental impacts. Ecological normalization should be individual for the NPP region of each ecosystem. The necessity to organize and conduct radiation ecological monitoring in the NPP regions is pointed out. Radiation ecological monitoring will provide for both environmental current radiation control and information for mathematical models, used in the NPPES radiation control

  10. Analysis for seismic response of dry storage facility for spent fuel

    International Nuclear Information System (INIS)

    Ko, Y.-Y.; Hsu, S.-Y.; Chen, C.-H.

    2009-01-01

    Most of the dry storage systems for spent fuel are freestanding, which leads to stability concerns in an earthquake. In this study, as a safety check, the ABAQUS/Explicit code is adopted to analyse the seismic response of the dry storage facility planned to be installed at Nuclear Power Plant no. 1 (NPP1) in Taiwan. A 3D coupled finite element (FE) model was established, which consisted of a freestanding cask, a concrete pad, and underneath soils interacting with frictional contact interfaces. The scenario earthquake used in the model included an artificial earthquake compatible to the design spectrum of NPP1, and a strong ground motion modified from the time history recorded during the Chi-Chi earthquake. The results show that the freestanding cask will slide, but not tip over, during strong earthquakes. The scale of the sliding is very small and a collision between casks will not occur. In addition, the differential settlement of the foundation pad that takes place due to the weight of the casks increases the sliding potential of the casks during earthquakes

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

    International Nuclear Information System (INIS)

    Bilanin, W.; Sills, S.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Mulyono Daryoko

    2007-01-01

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

  13. Russian normative approach to the question of management of NPP life time

    International Nuclear Information System (INIS)

    Karpunin, N.I.

    2002-01-01

    Full text: In Russia, the designated service life of a nuclear power plant (NPP) is 30 years. During the period 2001-2010, 15 Russian NPP units will reach the end of their service life. The 'Basic Provisions of NPP Safety Assurance', OPB-88/97, Item 5.1.14, provide for a possible extension of NPP operation beyond the designated service life. For such an extension, the NPP operating organization must apply for a license renewal to Gosatomnadzor, which needs to specify the relevant requirements. GAN is developing regulatory documents to provide a basis for NPP license renewal/extension of NPP operation, which would benefit from international experience. In accordance with 'The program of Atomic Energy Development in the Russian Federation for 1998-2005 and up to 2010' adopted by Decree No. 815 of the Government of the Russian Federation on 21 July 1998, priority is placed on the preparation of NPPs for extension of service life and on ensuring safety in the extended operating period. The length of the extension beyond the designated service life is to be determined on the basis of a range of technical and economic considerations, including: The ability to ensure and maintain operational safety; Sufficient residual service life of the unit's non-repairable components; The availability of temporary storage facilities for the additional spent nuclear fuel and radioactive waste, or the possibility of its transport off-site; The ability to ensure safe handling of the radioactive waste generated during the extension period; To extend the lifetime of an NPP unit, the plant Operator is required to perform the following tasks: Carry out a comprehensive survey of the NPP unit; Draw up a programme of preparation for lifetime extension; Prepare the NPP unit for operation in the extended period; Carry out the necessary tests. There are also some normative documents, which regulate management of NPP life time. (author)

  14. Technology for NPP decantate treatment realized at Kola NPP

    International Nuclear Information System (INIS)

    Stakhiv, Michael; Avezniyazov, Slava; Savkin, Alexander; Fedorov, Denis; Dmitriev, Sergei; Kornev, Vladimir

    2007-01-01

    At Moscow SIA 'Radon' jointly with JSC 'Alliance Gamma', the technology for NPP Decantate Treatment was developed, tested and realized at Kola NPP. This technology consists of dissolving the salt residue and subsequent treatment by ozonization, separation of the deposits formed from ozonization and selective cleaning by ferro-cyanide sorbents. The nonactive salt solution goes to an industrial waste disposal site or a repository specially developed at NPP sites for 'exempt waste' products by IAEA classification. This technology was realized at Kola NPP in December 2006 year. At this time more than 1000 m 3 of decantates log time stored are treated. It allows solving very old problem to empty decantates' tanks at NPPs in environmentally safe manner and with high volume reduction factor. (authors)

  15. A rational evaluation of structural design loads

    International Nuclear Information System (INIS)

    Tasaka, S.

    1993-01-01

    The reliability-based seismic design of structures is a design method ensuring that the structural seismic capacity is not less than the maximum seismic load or load effect for a prescribed value of the reliability index, wherein the design reference period, n, is used to specify the n-year maximum load. In the conventional Load and Resistance Factor Design (LRFD) method the design load is commonly determined on the basis of the n-year maximum the probability distribution of which may be given in some different ways. However, in contrast with the structural capacity the n-year maximum load usually involves much larger variabilities. The effort to decrease the variability would, hence, be effective for the purpose of avoiding nuclear power plant (NPP) structures having unnecessarily large capacities. A possible way to do this is to consider the joint probability distribution of the n-year 1st and 2nd maxima of the seismic load derived from the formula of extreme order statistics. Since the reliability index is conventionally associated with the n-year 1st maximum, the conditional probability distribution rather than the joint one of the n-year 1st maximum given a value of the n-year 2nd one will be considered. Three conditional extreme value distributions, which correspond to the usual extreme value distributions of Types I, II and III, and their statistical moments up to the second order are presented. Within the framework of the first-order second moment method, the conditional statistical moments are utilized to calculate the reliability index as well as the design value of the seismic load. The seismic load considered herein is represented by the peak ground acceleration (PGA) in n years. The present scheme is applied to evaluate the design PGA's at II sites in Japan where samples of the annual 1st and 2nd PGA's have been obtained by using historical seismic data. In this application the following two points are of our interest: (a) Define the reliability

  16. Seismic isolation of nuclear power plants using elastomeric bearings

    Science.gov (United States)

    Kumar, Manish

    Seismic isolation using low damping rubber (LDR) and lead-rubber (LR) bearings is a viable strategy for mitigating the effects of extreme earthquake shaking on safety-related nuclear structures. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The nuclear accident at Fukushima Daiichi in March 2011 has led the nuclear community to consider seismic isolation for new large light water and small modular reactors to withstand the effects of extreme earthquakes. The mechanical properties of LDR and LR bearings are not expected to change substantially in design basis shaking. However, under shaking more intense than design basis, the properties of the lead cores in lead-rubber bearings may degrade due to heating associated with energy dissipation, some bearings in an isolation system may experience net tension, and the compression and tension stiffness may be affected by the horizontal displacement of the isolation system. The effects of intra-earthquake changes in mechanical properties on the response of base-isolated nuclear power plants (NPPs) were investigated using an advanced numerical model of a lead-rubber bearing that has been verified and validated, and implemented in OpenSees and ABAQUS. A series of experiments were conducted at University at Buffalo to characterize the behavior of elastomeric bearings in tension. The test data was used to validate a phenomenological model of an elastomeric bearing in tension. The value of three times the shear modulus of rubber in elastomeric bearing was found to be a reasonable estimate of the cavitation stress of a bearing. The sequence of loading did not change the behavior of an elastomeric bearing under cyclic tension, and there was no

  17. Characterization of design ground motion for the central and eastern United States: licensing implications

    International Nuclear Information System (INIS)

    Litehiser, J.; Carrato, P.

    2005-01-01

    For the first time in decades several US utilities are exploring the possibility of building new Nuclear Power Plant (NPP) generating capacity in the Central and Eastern United States (CEUS). Among the many topics that must be considered to license a nuclear plant (NPP) is appropriate design to mitigate the potential effects of vibratory ground motion from earthquakes. Agreement on seismic design ground motion was not always easy during licensing of the last generation of NPPs. Therefore, over the last few decades both industry and the United States Nuclear Regulatory Commission (USNRC) have worked to find ground motion criteria that recognize and overcome earlier licensing difficulties. Such criteria should be stable and easily implemented. Important and complementary programs under the direction of the Lawrence Livermore National Laboratory (LLNL) and the Electric Power Research Institute (EPRI) were part of this effort, and these studies resulted in probabilistic seismic hazard assessments (PSHAs) for a number of CEUS NPP sites. These results and the concepts underlying them are now incorporated into both USNRC regulation and regulatory guidance. Nevertheless, as the utilities and the NRC begin a renewed licensing dialog, issues of regulatory interpretation of earthquake ground motion design criteria have emerged. These issues are as fundamental as the shape and amplitude of ground motion design response spectra and as significant as the impact of these spectra on structural design. Successful and timely resolution of these issues will significantly impact the future of nuclear power in the US. The purpose of this paper is to briefly describe some of these issues and the approaches that have been proposed for their resolution. (authors)

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

    Directory of Open Access Journals (Sweden)

    M. Budescu

    2009-01-01

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

  19. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

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

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

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

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

    Directory of Open Access Journals (Sweden)

    MOHD IRWAN ADIYANTO

    2014-02-01

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

  2. Decommissioning of Brennilis NPP

    International Nuclear Information System (INIS)

    Baize, Jean-Marc

    1998-01-01

    This EDF press communique give information related to the decommissioning of the Brennilis NPP. The following five items are developed in this report: 1. the level-2 decommissioning operations at the Brennilis NPP; 2. the Brennilis NPP, a pilot operation from the commissioning up to the decommissioning; 3. history of the Brennilis NPP decommissioning; 4. the types of radioactive wastes generated by the Brennilis NPP decommissioning; 5. the Brennilis NPP - a yard management as a function of the wastes. The document contains also seven appendices addressing the following subjects: 1. the share of decommissioning assigned to EDF and the decommissioning steps; 2. the EDF installations in course of decommissioning; 3. the CEA decommissioned installations or in course of decommissioning; 4. regulations; 5. costs; 6. waste management - principles; 7. data on the decommissioning yard

  3. Preliminary findings and lessons learned from the 16 July 2007 earthquake at Kashiwazaki-Kariwa NPP- 'The Niigataken Chuetsu-Oki earthquake, Kashiwazaki-Kariwa NPP and Tokyo, Japan, 6-10 August 2007. Mission report. V. 1

    International Nuclear Information System (INIS)

    2007-01-01

    Upon request from the Government of Japan an IAEA expert mission was conducted at the Kashiwazaki-Kariwa NPP following a strong earthquake that affected the plant on 16 July 2007. The objective, as agreed with the Japanese counterpart, was to conduct a fact finding mission and to identify the preliminary lessons learned that might have implications for the international nuclear safety regime. Although the Niigataken Chuetsu-Oki earthquake on 16 July 2007 significantly exceeded the level of the seismic input taken into account in the design of the plant, the installation behaved in a safe manner, during and after the earthquake. In particular, the automatic shutdown of the reactors of Units 3, 4 and 7, which were operating at full power, and of the reactor of Unit 2, which was in the start up state, were performed successfully. Based on the reports from experts from the Tokyo Electric Power Company (TEPCO) and the limited but representative plant walkdowns and visual observations performed by the IAEA team, safety related structures, systems and components of the plant seem to be in a much better general condition than might be expected for such a strong earthquake, and there is no visible significant damage. This is probably due to the conservatisms introduced at different stages of the design process. The combined effects of these conservatisms were apparently sufficient to compensate for uncertainties in the data and methods available at the time of the design of the plant, which led to the underestimation of the original seismic input. However, important components like the reactor vessels, the core internals and the fuel elements have not yet been examined and in-depth inspections are still to be performed. On the other hand, non-safety related structures, systems and components were affected by significant damage such as soil and anchorage failures and oil leakages. A re-evaluation of the seismic safety the Kashiwazika-Kariwa NPP needs to be done with account

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  6. NPP Krsko Living PSA Concept

    International Nuclear Information System (INIS)

    Vrbanic, I.; Spiler, J.

    2000-01-01

    NPP Krsko developed PSA model of internal and external initiators within the frame of the Individual Plant Examination (IPE) project. Within this project PSA model was used to examine the existing plant design features. In order to continue with use of this PSA model upon the completion of IPE in various risk-informed applications in support of plant operation and evaluations of design changes, an appropriate living PSA concept needed to be defined. The Living PSA concept is in NPP Krsko considered as being a set of activities pursued in order to update existing PSA model in a manner that it appropriately represents the plant design, operation practice and history. Only a PSA model which is being updated in this manner can serve as a platform for plant-specific risk informed applications. The NPP Krsko living PSA concept is based on the following major ponts. First, the baseline PSA model is defined, which is to be maintained and updated and which is to be reference point for any risk-informed application. Second, issues having a potential for impact on baseline PSA model are identified and procedure and responsibilities for their permanent monitoring and evaluation are established. Third, manner is defined in which consequential changes to baseline PSA model are implemented and controlled, together with associated responsibilities. Finally, the process is defined by which the existing version of baseline PSA model is superseded by a new one. Each time a new version of baseline PSA model is released, it would be re-quantified and the results evaluated and interpreted. By documenting these re-quantifications and evaluations of results in a sequence, the track is being kept of changes in long-term averaged risk perspective, represented by long-term averaged frequencies of core damage and pre-defined release categories. These major topics of NPP Krsko living PSA concept are presented and discussed in the paper. (author)

  7. Considerations related to Cernavoda NPP lifetime management

    International Nuclear Information System (INIS)

    Cojan, Mihail

    2007-01-01

    Cernavoda NPP, the first CANDU in Europe, is one of the original CANDU 6 plants and the first CANDU 6 producing over 706 MWe. While the first series of CANDU 6 plants (which entered service in the early 1980 s) have now reached the 2/3 of their 30 years design life, the Cernavoda NPP was put into service on the 2nd December 1996. After 10 years of operation the Plant Life Management (PLiM) Program for Cernavoda should be an increasingly important program to Utility ('CNE - Prod') in order to protect the investment and the continued success of plant operation. The goal of the paper is to present some considerations related to Cernavoda NPP lifetime management. The Plant Life Management Program, known as PLiM Program is concerned with the analysis of technical limits of the safe operation - from the point of view of nuclear safety - in NPP units, aiming at attaining the planned 30 years life duration and its extension to 40 or even 50 years of safe and economical operation. For the CANDU reactors the so-called PLiM and PLEX Programs are just applied. These are applied research programs that approach with priority the current practices for assessing the capability of safe operation within the limits of nuclear safety (fitness-for-service assessment). These programs also approach inspection, monitoring are prevention of degrading due to the ageing of critical systems, structures and components (CSSCs). As each nuclear plant is somewhat different in its components and systems, materials composition, procurement, construction, and operational history, directed research and development programs into materials behavior, monitoring techniques, and methods to mitigate ageing are required to support the lifetime management. Over the past 6 years, INR Pitesti (Institute for Nuclear Research - Romania) has been working on R and D Programs to support a comprehensive and integrated Cernavoda NPP Life Management Program (PLiM) that will see the Cernavoda NPP successfully and

  8. Decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    Anon

    2009-01-01

    In this presentation the Operation history of A1 NPP, Project 'Decommissioning of A1 NPP' - I stage, Project 'Decommissioning of A1 NPP ' - II stage and Next stages of Project 'Decommissioning of A1 NPP ' are discussed.

  9. Fusion of eastern and western technology in VVER 1000 NPP upgrade

    International Nuclear Information System (INIS)

    Ubra, O.; Fleischhans, J.; Kveton, M.

    1997-01-01

    An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

  10. Some remarks on the specification of seismic conditions for NPPs, based on the Romanian experience

    Energy Technology Data Exchange (ETDEWEB)

    Sandi, H [INCERC Building Research Inst., Bucharest (Romania)

    1993-07-01

    Some introductory remarks are devoted to the importance of revising the level of earthquake protection of NPPs, given the new knowledge at hand. Some summary data on the seismicity of Romania are given, with emphasis on the implications of instrumental data obtained during the recent strong intermediate depth earthquake of 1986 and 1990. Some problems raised by the new data in connection with the Cernavoda NPP (currently under construction), from the viewpoint of ground and floor design conditions, are discussed. Some proposals are then presented in connection with the future activities related to the verification of NPPs of the region. (author)

  11. Some remarks on the specification of seismic conditions for NPPs, based on the Romanian experience

    International Nuclear Information System (INIS)

    Sandi, H.

    1993-01-01

    Some introductory remarks are devoted to the importance of revising the level of earthquake protection of NPPs, given the new knowledge at hand. Some summary data on the seismicity of Romania are given, with emphasis on the implications of instrumental data obtained during the recent strong intermediate depth earthquake of 1986 and 1990. Some problems raised by the new data in connection with the Cernavoda NPP (currently under construction), from the viewpoint of ground and floor design conditions, are discussed. Some proposals are then presented in connection with the future activities related to the verification of NPPs of the region. (author)

  12. Temelin NPP commissioning experiences

    International Nuclear Information System (INIS)

    Hanus, V.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Test research and analysis for ultimate capacity of Qinshan NPP PCCV

    International Nuclear Information System (INIS)

    Zufeng, X.

    1994-01-01

    This paper introduces design and research for containment of Qinshan NPP which is the first PWR in CHINA designed and constructed by ourselves. The PCCV design is basically in conformity to ASME code. To verify the structural integrity capacity of Qinshan NPP containment, we fulfilled SIT and ILRT successfully in June, 1991. The special attention of the paper is focused on the ultimate capacity of the PCCV under severe accidents and earthquake. A study comprised of five different independent parts has been performed for the development of containment model test and corresponding nonlinear analysis. There are two prestressed concrete containment models with equipment hatch. One is 1/15 scale with steel liner tested on shake table and then moved out loaded with atmospheric pressure. The other is 1/10 scale without steel liner loaded with water pressure until destruction. From different methods including model test and nonlinear analysis, all obtained unanimous conclusion. The capacity under internal pressure and earthquake is reliable. The safety margin is enough. Consequently, in the second phase of Qinshan NPP and other PWR NPP under design, PCCV should be a better selection in China since it's more economic, rational and safe. (author)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  16. NPP Krsko on-line low pressure containment tightness monitoring implementation

    International Nuclear Information System (INIS)

    Dudas, M.; Basic, I.

    2004-01-01

    Containment Integrated Leak Rate Test (CILRT) 1999 in NPP Krsko was completely performed following regulation of 10CFR50 Appendix J Option A and ANSI/ANS 56.8-1987 at a design pressure (3.15 kp/cm2). In 2001 NPP Krsko proposed to Slovenian Nuclear Safety Administration (SNSA) the Technical Specification (TS) and Updated Safety Analysis Report (USAR) changes that describe implementation of new test intervals for Type A, B and C tests according to 10CFR50, Appendix J, Option B. After the positive final independent review of proposed changes by Authorized Institution, NPP Krsko received the License Amendment requiring from NPP Krsko to define technical solution for surveillance of containment tightness between two 10-years CILRT. This paper intends to discuss proposed methods by NPP Krsko, test equipment, performed measurements in 2004, associated analyses and evaluation.(author)

  17. Overview of Cernavoda NPP training programme

    International Nuclear Information System (INIS)

    Valache, Cornelia

    2005-01-01

    Full text: The paper presents a general assessment of Cernavoda NPP personnel training programme, highlighting the role of training in human performance improvement. Cernavoda NPP Personnel Training and Authorization Department (PTAD) is the one responsible for the training of Unit 1 and 2 staff. PTAD is structured in such a manner as to support and respond to all Cernavoda NPP training, qualification and authorization requirements. The training of personnel is continuously adapted based on IAEA Guides and INPO/WANO recommendations, to keep with world standards, based on the internal and external reviews. At Cernavoda NPP the Training Concept and the Training Programmes are based on SAT - Systematic Approach to Training. The Training Concept is established on a set of training documents (RD's, SI's, IDP's), which address all the SAT phases: Analysis, Design, Development, Implementation and Evaluation. The Training Programmes are structured on the initial and continuing personnel training. Their content and goals respond to the training specific needs for each plant major job family. In order to successfully support NPP training programmes, CNPP training center has upgraded classrooms with new presentation facilities and there are plans to expand the space of the building, to develop additional operator and maintenance skills facilities. There are also a number of general areas for improvement of PTAD activity, such as operators and maintenance training, initial and continuing training of both Unit 1 and 2 staff. By responding in a timely and completely manner to all plant training requirements PTAD will help in rising human performance of Cernavoda NPP personnel, supporting the safe, efficient and cost effective production of power. (author)

  18. 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)

  19. Preparation of Long Term Operation in Dukovany NPP, Czech Republic

    International Nuclear Information System (INIS)

    Krivanek, R.; Sabata, M.

    2012-01-01

    Dukovany NPP in the south-east of the Czech Republic operates four VVER 440/213 type units. The first unit was commissioned in 1985 and the last one in 1987. The operational results of the whole NPP have been excellent and NPP permanently belongs between the first quartile of the best operated NPPs in the world in accordance with WANO factors. Large safety improvement programme have been implemented in last 15 years. The original design lifetime of main components is 30 years which means till 2015 and it is understandable that NPP is preparing for long-term operation (LTO). The paper is describing activities carried out and planned for safe and successful LTO. (author)

  20. 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

  1. 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.

  2. Seismic design principles for the German fast breeder reactor SNR 2

    International Nuclear Information System (INIS)

    Busch, K.A.; Peters, K.A.; Rosenhauer, W.

    1987-01-01

    The safety issue of an adequate and optimized external event protection is of course that unnecessary hardware precautions might promote internal disturbances or hamper their control. It has up to now not satisfactorily been realized that the only serious context for seismic impacts on a fast reactor is their attributed potential of overriding core disruptive accident prevention, see e.g. GRS 1982. General and exaggerated antiseismic design features not focussed upon this point may as well turn out to be non-negligible initators in the absence of seismic vibrations. Unexpected snubber difficulties requiring additional reactor scrams and decay heat removal phases may be named as a simple example. The presented seismic design principles reflect the progress made in the concerned fields of analysis and do serve on the other hand as guidelines for research and development efforts under work. (orig./GL)

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

    International Nuclear Information System (INIS)

    Singh, N.; Duff, C.G.

    1979-10-01

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

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

    DEFF Research Database (Denmark)

    Katsanos, Evangelos; Vamvatsikos, Dimitrios

    2017-01-01

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

  5. Calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs shake table investigation at Paks NPP. Final report from 15 June 1993 - 15 June 1994

    International Nuclear Information System (INIS)

    Kaznovsky, S.

    1995-01-01

    This final report involves the calculation and experimental examination and ensuring the seismic resistance of the reactor equipment and pipelines at start up and operation of WWER type nuclear power plants. Shake table experiments performed at the Paks NPP are included. Namely the following devices of the emergency cooling system were tested: pump of low pressure; valve of low pressure; intermediate heat exchanger. The following values were determined: natural frequencies and vibration decrements and the main modes of normal vibrations for the heat exchanger

  6. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

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

    Science.gov (United States)

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

    2017-04-01

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

  8. Seismic analysis, evaluation and upgrade design for a nuclear facility exhaust stack building

    International Nuclear Information System (INIS)

    Malik, L.E.; Kabir, A.F.

    1991-01-01

    This paper reports on an exhaust stack building of a nuclear reactor facility with complex structural configuration that has been analyzed and evaluated for seismic forces. This building was built in the 1950's and had not been designed to resist seismic forces. A very rigorous analysis and evaluation program was implemented to minimize the costly retrofits required to upgrade the building to resist high seismic forces. The seismic evaluations were performed for the building in its as-is configuration, and as modified for several upgrade schemes. Soil-structure-interaction, base mat flexibility and the influence of the nearby reactor building have been considered in the seismic analyses. The rigorous analyses and evaluation enabled limited upgrades to qualify the stack building for the seismic forces

  9. Probabilistic safety analysis forecast for Trillo 1 NPP

    International Nuclear Information System (INIS)

    Carretero Fernandino, J.A.; Martin Alvarez, L.; gomez, F.; Cuallado, G.

    1995-01-01

    The performance of Probabilistic Safety Analyses (PSA) at Trillo 1 NPP is facing a number of challenges, unprecedented in previous PSAs carried out in Spain, due to the particular design characteristics of the plant. On account of this, it has been necessary to implemented specific approaches and methodological alternatives to perform a PSA which, while maintaining detail level and requirements in line with PSAs carried out previously in Spain, offers a solution technically adapted to the characteristics of the SIEMENS-KWU design as opposed to other Spanish reactors with a basic Westinghouse-General Electric design, which are based on standard US design. The purpose of this paper is to describe the most significant characteristics of the PSA at Trillo 1 NPP and the methodology used to date, taking into account current project progress

  10. Seismic design of circular-section concrete-lined underground openings: Preclosure performance considerations for the Yucca Mountain Site

    International Nuclear Information System (INIS)

    Richardson, A.M.; Blejwas, T.E.

    1992-01-01

    Yucca Mountain, the potential site of a repository for high-level radioactive waste, is situated in a region of natural and man-made seismicity. Underground openings excavated at this site must be designed for worker safety in the seismic environment anticipated for the preclosure period. This includes accesses developed for site characterization regardless of the ultimate outcome of the repository siting process. Experience with both civil and mining structures has shown that underground openings are much more resistant to seismic effects than surface structures, and that even severe dynamic strains can usually be accommodated with proper design. This paper discusses the design and performance of lined openings in the seismic environment of the potential site. The types and ranges of possible ground motions (seismic loads) are briefly discussed. Relevant historical records of underground opening performance during seismic loading are reviewed. Simple analytical methods of predicting liner performance under combined in situ, thermal, and seismic loading are presented, and results of calculations are discussed in the context of realistic performance requirements for concrete-lined openings for the preclosure period. Design features that will enhance liner stability and mitigate the impact of the potential seismic load are reviewed. The paper is limited to preclosure performance concerns involving worker safety because present decommissioning plans specify maintaining the option for liner removal at seal locations, thus decoupling liner design from repository postclosure performance issues

  11. Review of public comments on proposed seismic design criteria

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  12. Mochovce NPP simulator

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    Mochovce NPP simulator basic features and detailed description of its characteristics are presented with its performance, certification and application for training of NPP operators as well as the training scenario

  13. Earthquake response spectra for seismic design of nuclear power plants in the UK

    International Nuclear Information System (INIS)

    Bommer, Julian J.; Papaspiliou, Myrto; Price, Warren

    2011-01-01

    Highlights: → Seismic design of UK nuclear power plants usually based on PML response spectra. → We review derivation of PML spectra in terms of earthquake data used and procedure. → The data include errors and represent a small fraction of what is now available. → Seismic design loads in current practice are derived as mean uniform hazard spectra. → The need to capture epistemic uncertainty makes use of single equation indefensible. - Abstract: Earthquake actions for the seismic design of nuclear power plants in the United Kingdom are generally based on spectral shapes anchored to peak ground acceleration (PGA) values obtained from a single predictive equation. Both the spectra and the PGA prediction equation were derived in the 1980s. The technical bases for these formulations of seismic loading are now very dated if compared with the state-of-the-art in this field. Alternative spectral shapes are explored and the options, and the associated benefits and challenges, for generating uniform hazard response spectra instead of fixed shapes anchored to PGA are discussed.

  14. Operation experience of the advanced fuel assemblies at Unit 1 of Volgodonsk NPP within four fuel cycles

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Kobelev, S.; Kushmanov, S.

    2006-01-01

    The first commissioning of Volgodonsk NPP Unit 1 with standard reactor WWER-1000 (project V-320) was in 2001. The reactor core, starting from the first fuel charge, was arranged completely with Advanced Fuel Assemblies (AFAs). In this way, it is possible to obtain the experience in startup and operation of the core, completely arranged with AFAs, and also to get a possibility of performing the comprehensive check for justification of newly commissioned units and justification of design solutions accepted in the design of reactor core for Taiwan NPP, Bushehr NPP and Kudankulam NPP. The first fuel charge of the Volgodonsk NPP Unit 1 is a reference and unified for Tiawan NPP (V-428), Bushehr NPP (V-446), Kudankulam NPP(V-412) with small differences caused by design features of RP V-320. The first core charge of Unit 1 of Volgodonsk NPP was arranged of 163 AFAs, comprising 61 CPS ARs and 42 BAR bundles. The subsequent fuel charges were arranged of AFAs with gadolinium oxide integrated into fuel instead of BAR. By 2005 the results of operation of the core at Unit 1 of Volgodonsk NPP during four fuel cycles showed that AFA is sufficiently reliable and serviceable. The activity of the primary coolant of the Volgodonsk NPP is at stable low level. During the whole time of the core operation of the Volgodonsk NPP Unit 1 no leaky AFAs were revealed. The modifications of the internals, made during pre-operational work, are reasonable and effective to provide for fuel mechanical stability in the course of operation. The modifications, made in AFA structure during operation of the Volgodonsk NPP Unit 1, are aimed at improving the service and operational reliability of its components. Correctness of the solutions taken is confirmed by AFAs operation experience both at the Volgodonsk NPP, and at other operating Russian NPPs

  15. Failure Capacity Evaluation for Anchor System of NPP Facilities by using a Shaking Table Test

    International Nuclear Information System (INIS)

    Kwon, Hyung O; Jung, Min Ki; Park, Jin Wan; Lim, Ji Hoon

    2010-02-01

    This study investigate the destructive influence of crack locations on the anchor performance to evaluate the seismic performance of NPP equipment anchored on damaged concrete. For this purpose, small-scale specimens were fabricated according to the following three cases: 1) with a non-damaged anchor; 2) with cracks running through the anchor; and 3) with cracks along the expected corn-shape fracture away from the anchor. The result verified with the finite element method is as follows: In the first and second cases that is, with a non-damaged anchor and with cracks running through the anchor destruction occurred at the anchor steel. In the third case that is, with cracks around the anchor, a 30% decline in the seismic performance was identified. This result indicates that evaluation of seismic performance and relevant reinforcement are required when cracks occur away from the anchor along the expected destructive surface

  16. Two managerial grids in NPP

    International Nuclear Information System (INIS)

    Zhao Hui

    2012-01-01

    Today, the nuclear power corporation (NPC) enjoys the profit of LCEP (the low carbon economic policy). at the same time, they also enduring more and more pressure. For example, the partner competition or the NPP potential occupational risk . The efficient counterplot of risk is the self-ability cultivation. It is essential to research the NPP managerial flow. The nuclear power plant (NPP) unit is a carrier of the NPC enterprise management system, and has taken on a new look 'pull one portion then the whole moving'. The NPP has three systematical characters, the security responsibility center, the man-machine system and the input-output system. The manufacturing system and the enterprise management system are the great constituents of the NPP managerial flows. Means of systems analysis, we can find out the truth of the NPP running interface. In CHINA, there are many operating experiences near 20 years. It indicates that the NPP manufacturing system and the enterprise system are the roots of the nuclear power corporation, the core of the all NPP systems must be based on it. So the ability cultivation is the work core to NPP. It is reliably to ensure the NPP to be up against problems, for instance, the security duty, the costing control and the man-machine system running harmoniously. This paper introduces the NPP managerial flow and the present state of QNPC, also come up with a proposal to refer for the NPC development actions of collective measure, specialization, standardization, fine. (author)

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

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2004-01-01

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

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

    Science.gov (United States)

    Glezil, Dorothy

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

  19. 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

  20. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  1. Preclosure seismic design methodology for a geologic repository at Yucca Mountain. Revision 1

    International Nuclear Information System (INIS)

    1996-08-01

    This topical report is the second in a series of three reports being developed by the US Department of Energy (DOE) to document the preclosure seismic design of structures, systems, and components (SSCs) that are important to the radiological safety of the potential repository at Yucca Mountain, Nevada. The first topical report, Methodology to Assess Fault Displacement and Vibratory Ground Motion Hazards at Yucca Mountain, YMP/TR-002-NP, was submitted to the US Nuclear Regulatory Commission (NRC) staff for review and comment in 1994 and has been accepted by the staff. The DOE plans to implement this methodology in fiscal year 1997 to develop probabilistic descriptions of the vibratory ground motion hazard and the fault displacement hazard at the Yucca Mountain site. The second topical report (this report) describes the DOE methodology and acceptance criteria for the preclosure seismic design of SSCs important to safety. A third report, scheduled for fiscal year 1998, will document the results of the probabilistic seismic hazard assessment (conducted using the methodology in the first topical report) and the development of the preclosure seismic design inputs. This third report will be submitted to NRC staff for review and comment as a third topical report or as a design study report

  2. SEISMIC DESIGN REQUIREMENTS SELECTION METHODOLOGY FOR THE SLUDGE TREATMENT and M-91 SOLID WASTE PROCESSING FACILITIES PROJECTS

    International Nuclear Information System (INIS)

    RYAN GW

    2008-01-01

    In complying with direction from the U.S. Department of Energy (DOE), Richland Operations Office (RL) (07-KBC-0055, 'Direction Associated with Implementation of DOE-STD-1189 for the Sludge Treatment Project,' and 08-SED-0063, 'RL Action on the Safety Design Strategy (SDS) for Obtaining Additional Solid Waste Processing Capabilities (M-91 Project) and Use of Draft DOE-STD-I 189-YR'), it has been determined that the seismic design requirements currently in the Project Hanford Management Contract (PHMC) will be modified by DOE-STD-1189, Integration of Safety into the Design Process (March 2007 draft), for these two key PHMC projects. Seismic design requirements for other PHMC facilities and projects will remain unchanged. Considering the current early Critical Decision (CD) phases of both the Sludge Treatment Project (STP) and the Solid Waste Processing Facilities (M-91) Project and a strong intent to avoid potentially costly re-work of both engineering and nuclear safety analyses, this document describes how Fluor Hanford, Inc. (FH) will maintain compliance with the PHMC by considering both the current seismic standards referenced by DOE 0 420.1 B, Facility Safety, and draft DOE-STD-1189 (i.e., ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, and ANSI ANS 2.26-2004, Categorization of Nuclear Facility Structures, Systems and Components for Seismic Design, as modified by draft DOE-STD-1189) to choose the criteria that will result in the most conservative seismic design categorization and engineering design. Following the process described in this document will result in a conservative seismic design categorization and design products. This approach is expected to resolve discrepancies between the existing and new requirements and reduce the risk that project designs and analyses will require revision when the draft DOE-STD-1189 is finalized

  3. 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)

  4. Changing NPP consumption patterns in the Holocene: from Megafauna "liberated" NPP to "ecological bankruptcy"

    Science.gov (United States)

    Doughty, C.

    2015-12-01

    There have been vast changes in how net primary production (NPP) is consumed by humans and animals during the Holocene beginning with a potential increase in availability following the Pleistocene megafauna extinctions. This was followed by the development of agriculture which began to gradually restrict availability of NPP for wild animals. Finally, humans entered the industrial era using non-plant based energies to power societies. Here I ask the following questions about these three energy transitions: 1. How much NPP energy may have become available following the megafauna extinctions? 2. When did humans, through agriculture and domestic animals, consume more NPP than wild mammals in each country? 3. When did humans and wild mammals use more energy than was available in total NPP in each country? To answer this last question I calculate NPP consumed by wild animals, crops, livestock, and energy use (all converted to units of MJ) and compare this with the total potential NPP (also in MJ) for each country. We develop the term "ecological bankruptcy" to refer to the level of consumption where not all energy needs can be met by the country's NPP. Currently, 82 countries and a net population of 5.4 billion are in the state of ecologically bankruptcy, crossing this threshold at various times over the past 40 years. By contrast, only 52 countries with a net population of 1.2 billion remain ecologically solvent. Overall, the Holocene has seen remarkable changes in consumption patterns of NPP, passing through three distinct phases. Humans began in a world where there was 1.6-4.1% unclaimed NPP to consume. From 1700-1850, humans began to consume more than wild animals (globally averaged). At present, >82% of people live in countries where not even all available plant matter could satisfy our energy demands.

  5. Kozloduy Nuclear Power Plant (Unit 1 and 2). Proposed modifications to increase the seismic capability of equipment and main structures

    International Nuclear Information System (INIS)

    Ordonez Villalobos, A.; Monette, P.R.

    1993-01-01

    Within the framework of the European Community's PHARE Programme of improvement to facilities, their operating systems, equipment and buildings of the Kozloduy NPP in Bulgaria, plant safety during seismic events is considered to be an issue of overriding importance, especially in view of the earthquakes the region suffered during the last decade. Westinghouse Energy Systems International (WESI) and Empresarios Agrupados (EA) have initiated an intensive programme for physical upgrading of equipment with a view to augmenting its seismic capability and, at the same time, to studying design modifications in the diesel-generator buildings, pump house and main building structures (turbines, electrical building). The implementation of these modifications requires an in situ inspection of the real conditions of the various elements, analyses, conceptual design and detail engineering, all of which has to be done in short periods of time using resources available at the plant. This activity is performed by the companies mentioned above, with the collaboration of two engineering companies, Energoproekt of Bulgaria and INITEC of Spain. This paper describes the activities developed and the treatment given to the various aspects of improvement of the seismic capability of equipment and structures. (author)

  6. 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)

  7. The 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1995-01-01

    A record of the 1994 Forum on Appropriate Criteria and Methods for Seismic Design of Nuclear Piping is provided. The focus of the forum was the design-by-rule method for seismic design of piping. Issues such as acceptance criteria, ductility considerations, demonstration of margin, training, verification and costs were discussed. The use of earthquake experience data, including the recent Northridge earthquake, to justify a design-by-rule method was explored. The majority of the participants felt there are not significant advantages to developing a design-by-rule approach for new plant design. One major disadvantage was considered by many to be training. Extensive training will be required to properly implement a design-by-rule approach. Verification of designs was considered by the majority to be equally important for design-by-rule as for design-by-analysis. If a design-by-rule method is going to be effective, the method will have to be based on ductility considerations (UBC approach). A significant issue will be justification of seismic margins with liberal rules. The UBC approach is being questioned by some because of the recent structural cracking problems in the Northridge earthquake

  8. Seismic isolation of nuclear power plants using sliding isolation bearings

    Science.gov (United States)

    Kumar, Manish

    Nuclear power plants (NPP) are designed for earthquake shaking with very long return periods. Seismic isolation is a viable strategy to protect NPPs from extreme earthquake shaking because it filters a significant fraction of earthquake input energy. This study addresses the seismic isolation of NPPs using sliding bearings, with a focus on the single concave Friction Pendulum(TM) (FP) bearing. Friction at the sliding surface of an FP bearing changes continuously during an earthquake as a function of sliding velocity, axial pressure and temperature at the sliding surface. The temperature at the sliding surface, in turn, is a function of the histories of coefficient of friction, sliding velocity and axial pressure, and the travel path of the slider. A simple model to describe the complex interdependence of the coefficient of friction, axial pressure, sliding velocity and temperature at the sliding surface is proposed, and then verified and validated. Seismic hazard for a seismically isolated nuclear power plant is defined in the United States using a uniform hazard response spectrum (UHRS) at mean annual frequencies of exceedance (MAFE) of 10-4 and 10 -5. A key design parameter is the clearance to the hard stop (CHS), which is influenced substantially by the definition of the seismic hazard. Four alternate representations of seismic hazard are studied, which incorporate different variabilities and uncertainties. Response-history analyses performed on single FP-bearing isolation systems using ground motions consistent with the four representations at the two shaking levels indicate that the CHS is influenced primarily by whether the observed difference between the two horizontal components of ground motions in a given set is accounted for. The UHRS at the MAFE of 10-4 is increased by a design factor (≥ 1) for conventional (fixed base) nuclear structure to achieve a target annual frequency of unacceptable performance. Risk oriented calculations are performed for

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. New Emergency control center of Slovenske Elektrarne Bohunice NPP

    International Nuclear Information System (INIS)

    Pecko, E.

    2012-01-01

    Emergency preparedness of nuclear power plant and operation assurance in case of a possible emergency calls to have devices for emergency response with equipment for rapid detection and continuous evaluation of anticipated events. Chief executive body designated to manage a nuclear power plant during major events is the emergency committee (EC). Emergency Committee is a part of the Emergency Response Organization (ERO). The following centers are on alert to ensure the activities of the ERO - Emergency Response office: - control room (CR) and emergency control room (ECR); - emergency management center (EMC); - Monitoring Centre (MC); - emergency backup control center (EBCC); - congregations of civil protection (CP) and CP shelters; - communications with warning and notification system (VARVYR). From a historical and practical point of view in the vicinity of Jaslovske Bohunice has been set up a joint emergency control center. The center was located on the territory of the former already inoperative V1 NPP. V1 NPP is currently integrated into the organizational structure of JAVYS. Operating Bohunice V2 NPP plant is a designated part of the Slovenske Elektrarne, a. s., whose majority owner is an Italian operator ENEL. In terms of various relevant factors, it was decided to build a new emergency management center on the territory of operating V2 NPP, meet the current standards.

  11. Soil-structure interaction effects in seismic analysis of turbine generator building on rock-like foundation

    International Nuclear Information System (INIS)

    Park, Chi Seon; Lee, Sang Hoon; Yoo, Kwang Hoon

    2004-01-01

    Soil properties supporting structure may become criteria determining methodologies for seismic response analysis of a structure. Regulatory Guide describes that a fixed-base assumption is acceptable for structures supported on rock or rock-like materials defined by a shear wave velocity of 3,500 ft/sec or greater at a shear strain of 10 -3 percent or smaller when considering preloaded soil conditions due to the structure. Seismic analyses for the Korean nuclear power plant (NPP) structures satisfying the above site soil condition have been completed through the fixed-base analysis. However, dynamic responses for relatively stiff structures such as NPP structures still have soil-structure interaction (SSI) effects. In other words, the fixed-base analysis does not always yield conservative results to be compared with SSI analysis. The SSI effects due to different stiff soil properties for Turbine Generator Building (TGB) structure to be constructed at Kori site of South Korea are investigated in views of floor response spectra (FRS) and member forces

  12. Seismic stability of NPPs in Eastern Europe. Final report from June 1993 - June 1994

    International Nuclear Information System (INIS)

    Ambriashvili, Y.K.

    1995-01-01

    The present work was fulfilled within the contract between IAEA and 'Atomenegroproject' Institute of the Russian Federation. The program of the work fulfilled within the period from June 1993 till June 1994 was as follows: review and comments of the standards related to aseismic design of the structure and components; aseismic classification of main safety systems and components by function of WWER NPPs; response spectrum which was developed for design of WWER NPPs; comparison of Western and the Russian standards and codes related to aseismic design of NPPs. The general requirements for aseismic designed constructions, comments on the main important points of seismic loads, calculation of the seismic loads, methodology of design and test of the equipment and electrical systems are presented in the work. The main constructions and equipment for NPP WWER-440 provided the functions of plant during and after the earthquakes are also listed. Methodology of calculation and selection of response spectrum for design of WWER NPPs. The comparative analysis of the Western and the Russian Norms are reviewed. The main conclusions of the present work are as follows: General principal stages of the methods and approaches of the design, investigations and testing are almost identified with IAEA and Western approaches. Recommendations corresponding with the IAEA and Western standards should be consumed in a new version of codes of aseismic design of NPPs that are working out now in the Russian Federation. It is connected with maximum acceleration, response spectra, time histories, damping factor and etc. Meanwhile 'Atomenergoproject' now has possibility to fulfil calculations of soil- structure-equipment systems and carry out full scale tests of the equipment on the components seismic platform in accordance with the high level of the technical requirements

  13. Seismic stability of NPPs in Eastern Europe. Final report from June 1993 - June 1994

    Energy Technology Data Exchange (ETDEWEB)

    Ambriashvili, Y K [Atomenergoproject, Moscow (Russian Federation)

    1995-07-01

    The present work was fulfilled within the contract between IAEA and 'Atomenegroproject' Institute of the Russian Federation. The program of the work fulfilled within the period from June 1993 till June 1994 was as follows: review and comments of the standards related to aseismic design of the structure and components; aseismic classification of main safety systems and components by function of WWER NPPs; response spectrum which was developed for design of WWER NPPs; comparison of Western and the Russian standards and codes related to aseismic design of NPPs. The general requirements for aseismic designed constructions, comments on the main important points of seismic loads, calculation of the seismic loads, methodology of design and test of the equipment and electrical systems are presented in the work. The main constructions and equipment for NPP WWER-440 provided the functions of plant during and after the earthquakes are also listed. Methodology of calculation and selection of response spectrum for design of WWER NPPs. The comparative analysis of the Western and the Russian Norms are reviewed. The main conclusions of the present work are as follows: General principal stages of the methods and approaches of the design, investigations and testing are almost identified with IAEA and Western approaches. Recommendations corresponding with the IAEA and Western standards should be consumed in a new version of codes of aseismic design of NPPs that are working out now in the Russian Federation. It is connected with maximum acceleration, response spectra, time histories, damping factor and etc. Meanwhile 'Atomenergoproject' now has possibility to fulfil calculations of soil- structure-equipment systems and carry out full scale tests of the equipment on the components seismic platform in accordance with the high level of the technical requirements.

  14. NPOESS Preparatory Project (NPP) Science Overview

    Science.gov (United States)

    Butler, James J.

    2011-01-01

    NPP Instruments are: (1) well understood thanks to instrument comprehensive test, characterization and calibration programs. (2) Government team ready for October 25 launch followed by instrument activation and Intensive Calibration/Validation (ICV). NPP Data Products preliminary work includes: (1) JPSS Center for Satellite Applications and Research (STAR) team ready to support NPP ICV and operational data products. (2) NASA NPP science team ready to support NPP ICV and EOS data continuity.

  15. Seismic design of steel moment resisting frames-European versus American practice

    International Nuclear Information System (INIS)

    Naqash, M.T.; Matteis, G.D.; Luca, A.D.

    2012-01-01

    This paper provides an overview on the design philosophy of moment resisting frames (MRF) according to the seismic provisions of Eurocode 8 and American Institute of Steel Construction (AISC). A synopsis of the main recommendations of the two codes is briefly described. Then in order to examine the structural efficiency of the design principles of MRF according to the aforementioned codes, a case study is developed in which spatial and perimeter moment resisting frames of 12, 6 and 3 storeys residential building are considered. In the case of EC8, Ductility Class Medium (DCM) with behaviour factor of 4 and Ductility Class High (DCH) with behaviour factor of 6.5 for 6-storey frames are used, while only DCH is employed in the design of 12 and 3 storey frames. When dealing with AISC/American Society of Civil Engineers (ASCE) code, special moment resisting frame (SMF) with response modification factor of 8 is employed in the design. The outcomes from the design are illustrated in terms of frame performance, section profiles, strength-demand to capacity ratios, drift-demand to capacity ratios and structural weight, thus allowing the understanding of pros and cons of the design criteria and the capacity design rules of the two codes. The main purpose of the current paper is to compare the seismic design rules of the two codes with a parametric analysis developed by a case study in order to let the technician knows about the importance and influence of some important parameters which are given in the capacity design rules of the two codes. This study will be a benchmark for further analysis on the two codes for seismic design of steel structures. (author)

  16. The numerical computation of seismic fragility of base-isolated Nuclear Power Plants buildings

    International Nuclear Information System (INIS)

    Perotti, Federico; Domaneschi, Marco; De Grandis, Silvia

    2013-01-01

    Highlights: • Seismic fragility of structural components in base isolated NPP is computed. • Dynamic integration, Response Surface, FORM and Monte Carlo Simulation are adopted. • Refined approach for modeling the non-linearities behavior of isolators is proposed. • Beyond-design conditions are addressed. • The preliminary design of the isolated IRIS is the application of the procedure. -- Abstract: The research work here described is devoted to the development of a numerical procedure for the computation of seismic fragilities for equipment and structural components in Nuclear Power Plants; in particular, reference is made, in the present paper, to the case of isolated buildings. The proposed procedure for fragility computation makes use of the Response Surface Methodology to model the influence of the random variables on the dynamic response. To account for stochastic loading, the latter is computed by means of a simulation procedure. Given the Response Surface, the Monte Carlo method is used to compute the failure probability. The procedure is here applied to the preliminary design of the Nuclear Power Plant reactor building within the International Reactor Innovative and Secure international project; the building is equipped with a base isolation system based on the introduction of High Damping Rubber Bearing elements showing a markedly non linear mechanical behavior. The fragility analysis is performed assuming that the isolation devices become the critical elements in terms of seismic risk and that, once base-isolation is introduced, the dynamic behavior of the building can be captured by low-dimensional numerical models

  17. Chernobyl NPP decommissioning efforts - Past, Present and Future. Decommissioning Efforts on Chernobyl NPP site - Past, Present

    International Nuclear Information System (INIS)

    Kuchinskiy, V.

    2017-01-01

    Two unique large-scale projects are underway at the moment within the Chernobyl - Exclusion zone - Shelter object transformation into ecologically safe system and the decommissioning of 3 Chernobyl NPP Units. As a result of beyond design accident in 1986 the entire territory of the industrial site and facilities located on it was heavily contaminated. Priority measures were carried out at the damaged Unit under very difficult conditions to reduce the accident consequences and works to ensure nuclear and radiation safety are continuous, and the Unit four in 1986 was transformed into the Shelter object. Currently, works at the Shelter object are in progress. Under assistance of the International Community new protective construction was built above the existing Shelter object - New Safe Confinement, which will ensure the SO Safety for the long term - within up to 100 years. The second major project is the simultaneous decommissioning of Chernobyl NPP Units 1, 2 and 3. Currently existing Chernobyl NPP decommissioning Strategy has been continuously improved starting from the Concept of 1992. Over the years the following was analyzed and taken into account: the results of numerous research and development works, international experience in decommissioning, IAEA recommendations, comments and suggestions from the governmental and regulatory bodies in the fields of nuclear energy use and radioactive waste management. In 2008 the final decommissioning strategy option for Chernobyl NPP was approved, that was deferred gradual dismantling (SAFSTOR). In accordance with this strategy, decommissioning will be carried out in 3 stages (Final Shutdown and Preservation, Safe Enclosure, Dismantling). The SAFSTOR strategy stipulates: -) the preservation of the reactor, the primary circuit and the reactor compartment equipment; -) the dismantling of the equipment external in relation to the reactor; -) the safe enclosure (under the supervision); -) the gradual dismantling of the primary

  18. Radiological impact on the surrounding area of Cernavoda NPP operation

    International Nuclear Information System (INIS)

    Busnita, M.; Penescu, M.; Neacsu, L.; Ion, M.; Moldoveanu, E.

    1996-01-01

    The objective of this paper is to present the protective measures provided to Cernavoda NPP for population and environment in case of normal operation and accidental situations. The paper describes the nuclear safety concept with the two aspects, the technical safety and the radiation protection, respectively, which represent the theoretical basis of the NPP design, construction and operation. The assessment of the potential radiological effects of Cernavoda NPP operation on the population and other environmental factors have been done by using the following input data: the natural and social characteristics of the site and the technical characteristics of the plant. The effects of Cernavoda NPP operation are also exposed in health risks which are compared with the daily risks of the population exposed of the natural phenomena or to the activities like transportation, domestics or industrial. (author). 1 tab., 6 refs

  19. Transition cycle fuel management problems of NPP Krsko; Problemi gospodarenje gorivom u prijelaznim ciklusima NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia); Pevec, D [Elektrotehnicki fakultet, Zagreb (Yugoslavia); Smuc, T; Urli, N [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia)

    1989-07-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  20. The main trends of work on lifetime management of NPP

    International Nuclear Information System (INIS)

    Dragunov, Yu.; Kurakov, Yu.

    1995-01-01

    The major trends of works on NPP lifetime management are discussed: Co-ordinating and organizational activities; studying the degradation processes and residual life; elaboration and implementation of the measures to maintain/enhance safety level of NPPs operating within design life; development of control principals of advanced NPPs aging and safe life; creation and development of regulations in the field of NPP aging and life management. 7 figs

  1. Modernization and safety improvement project of the NPP V-2 Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Michal, V.; Losonsky, B.; Magdolen, J.

    2000-01-01

    This contribution deals with the form, present state, and results of the Nuclear Power Plants Research Institute (the Slovak acronym is VUJE - Vyskumny Ustav Jadrovych Elektrarni) participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project. A short description of VUJE history, activity, and results is also presented as well as NPPs Jaslovske Bohunice characterization. VUJE was established in 1977 and deals with scientific and research needs of nuclear power plants, such as design, construction, commissioning and operation. The next fields of VUJE activity are, NPP reconstruction, NPP personnel training, radioactive waste management technology, and NPP decommissioning. The nuclear power plant, Jaslovske Bohunice, is situated approximately 15 km from the district town of Trnava in the southwestern region of the Slovak Republic. The construction of the first Czechoslovak NPP A-1 began on this site in 1957 .The construction of the double-unit NPP V-1 with WWER-440 (type V-230) reactor began in 1972. The first unit of NPP V-1 began operation in 1978 and the second in 1980. NPPs construction on the Bohunice site continued with NPP V-2, which has two units with WWER-440 (type V-213) reactors. Unit 1 and Unit 2 of NPP V-2 were commissioned in 1984 and 1985, respectively. Slovak electric utility Slovenske elektrarne (SE) is the owner/commissioner of NPP V-2. This NPP is responsible for more than 20% of the total electrical energy production of SE, making it an essential supporter of the Slovak economy. (authors)

  2. NPP life management (abstracts)

    International Nuclear Information System (INIS)

    Litvinskij, L.L.; Barbashev, S.V.

    2002-01-01

    Abstracts of the papers presented at the International conference of the Ukrainian Nuclear Society 'NPP Life Management'. The following problems are considered: modernization of the NPP; NPP life management; waste and spent nuclear fuel management; decommissioning issues; control systems (including radiation and ecological control systems); information and control systems; legal and regulatory framework. State nuclear regulatory control; PR in nuclear power; training of personnel; economics of nuclear power engineering

  3. Preliminary findings and lessons learned from the 16 July 2007 earthquake at Kashiwazaki-Kariwa NPP- 'The Niigataken Chuetsu-Oki earthquake, Kashiwazaki-Kariwa NPP and Tokyo, Japan, 6-10 August 2007. Mission report. V. 1[Report to the Government of Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-08-17

    Upon request from the Government of Japan an IAEA expert mission was conducted at the Kashiwazaki-Kariwa NPP following a strong earthquake that affected the plant on 16 July 2007. The objective, as agreed with the Japanese counterpart, was to conduct a fact finding mission and to identify the preliminary lessons learned that might have implications for the international nuclear safety regime. Although the Niigataken Chuetsu-Oki earthquake on 16 July 2007 significantly exceeded the level of the seismic input taken into account in the design of the plant, the installation behaved in a safe manner, during and after the earthquake. In particular, the automatic shutdown of the reactors of Units 3, 4 and 7, which were operating at full power, and of the reactor of Unit 2, which was in the start up state, were performed successfully. Based on the reports from experts from the Tokyo Electric Power Company (TEPCO) and the limited but representative plant walkdowns and visual observations performed by the IAEA team, safety related structures, systems and components of the plant seem to be in a much better general condition than might be expected for such a strong earthquake, and there is no visible significant damage. This is probably due to the conservatisms introduced at different stages of the design process. The combined effects of these conservatisms were apparently sufficient to compensate for uncertainties in the data and methods available at the time of the design of the plant, which led to the underestimation of the original seismic input. However, important components like the reactor vessels, the core internals and the fuel elements have not yet been examined and in-depth inspections are still to be performed. On the other hand, non-safety related structures, systems and components were affected by significant damage such as soil and anchorage failures and oil leakages. A re-evaluation of the seismic safety the Kashiwazika-Kariwa NPP needs to be done with account

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Science.gov (United States)

    Tang, Hongliang

    2018-04-01

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

  6. Forecasting of resonances vibration equipment with elastic waves coolant and with the external periodic loads on NPP with WWER

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Zaporozhets, M.V.; Fedorov, A.I.

    2015-01-01

    Forecasting are carried out for external loads in relation to the main circulation circuit - dynamic loads caused by the rotation of the MCP, dynamic loads caused by the earthquake, dynamic loads caused by damage to the MCP in the earthquake. A comparison of the response spectrum of one of the variants of the base of the NPP, with the frequency vibration of the primary circuit equipment for NPP with WWER-1000 and self-frequency of elastic waves in the fluid. Analysis of the comparison results shows that the frequency of vibration of the main equipment of the reactor plant and elastic waves are in the frequency band in the spectrum response corresponding to the maximum amplitude of the seismic action [ru

  7. 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.)

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

    International Nuclear Information System (INIS)

    Wium, D.J.W.

    1997-01-01

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

  9. Fast Bayesian optimal experimental design for seismic source inversion

    KAUST Repository

    Long, Quan

    2015-07-01

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

  10. Fast Bayesian Optimal Experimental Design for Seismic Source Inversion

    KAUST Repository

    Long, Quan

    2016-01-06

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

  11. Fast Bayesian Optimal Experimental Design for Seismic Source Inversion

    KAUST Repository

    Long, Quan; Motamed, Mohammad; Tempone, Raul

    2016-01-01

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

  12. Studying the sampling representativeness in the NPP ventilation ducts

    International Nuclear Information System (INIS)

    Sosnovskij, R.I.; Fedchenko, T.K.; Minin, S.A.

    2000-01-01

    Measurements of the gas and aerosol voluminous activity in the NPP ventilation ducts are an important source of information on the radiation contaminants ingress into the environmental medium. These measurements include sampling, samples transport and proper measurements. The work is devoted to calculation of metrological characteristics of the sampling systems for the NPP gas-aerosol releases by different parameters of these systems and ventilation ducts. The results obtained are intended for application by designing such systems and their metrological certification [ru

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

  14. Numerical Investigation of Progressive Collapse Resistance for Seismically Designed RC Buildings

    OpenAIRE

    Marchiş, Adrian G.; Ioani, Adrian M.

    2014-01-01

    In this paper the progressive collapse behavior of a reinforced concrete framed building located in different seismic areas from Romania is investigated. The six-storey structure is designed for low (ag = 0.08 g), moderate (ag = 0.16 g) and high (ag = 0.24 g) seismic zone. Based on the GSA (2003) criteria, a nonlinear static analysis is conducted first in order to estimate the progressive collapse resistance of the models. It was shown that all the structures will collapse when subjected to i...

  15. NPP Temelin. Status of safety improvements

    International Nuclear Information System (INIS)

    1999-01-01

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

  16. NPP Evaluation, backfitting and life extension. An engineering viewpoint

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez Lopez, A [Empresarios Agrupados, A.I.E., Madrid (Spain)

    1993-12-15

    During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

  17. NPP Evaluation, backfitting and life extension. An engineering viewpoint

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.

    1993-01-01

    During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

  18. Safety improvement and results of commissioning of Mochovce NPP WWER 440/213

    International Nuclear Information System (INIS)

    Lipar, M.

    1998-01-01

    Mochovce NPP is the last one of this kind and compared to its predecessors, it is characterized by several modifications which contribute to the improvement of the safety level. In addition based on Nuclear Regulatory Authority requirements and based on documents: - IAEA - Safety Issues and their ranking for NPP WWER 440/213, - IAEA - Safety Improvement of Mochovce NPP Project Review Mission, - Riskaudit - Evaluation of the Mochovce NPP Safety Improvements. Additional safety measures have been implemented before commissioning. The consortium EUCOM (FRAMATOME - SIEMENS), SKODA Praha, ENERGOPROJEKT Praha, Russian organizations and VUJE Trnava Nuclear Power Plants research institute were selected for design and implementation of the safety measures. The papers summarized, safety requirements, safety measures implemented, results of commissioning and results of safety analysis report evaluation. (author)

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

    International Nuclear Information System (INIS)

    Novak, J.; Pecinka, L.

    1981-01-01

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

  20. Review of accident analyses performed at Mochovce NPP

    International Nuclear Information System (INIS)

    Siko, D.

    2000-01-01

    In this paper the review of accident analysis performed in NPP Mochovce V-1 is presented. The scope of these safety measures was defined and development in the T SSM for NPP Mochovce Nuclear Safety Improvements Report' issued in July 1995. The main objectives of these safety measures were the followings: (a) to establish the criteria for selection and classification of accidental events, as well as defining the list of initiating events to be analysed. Accident classification to the individual groups must be performed in accordance with RG 1.70 and IAEA recommendations 'Guidelines for Accidental Analysis of WWER NPP' (IAEA-EBR-WWER-01) to select boundary cases to be calculated from the scope of initiating events; (b ) to elaborate the accident analysis methodology that also includes acceptance criteria for their result evaluation, initial and boundary conditions, assumption related with the application of the single failure criteria, requirements on the analysis quality, used computer codes, as well as NPP models and input data for the accident analysis; (c) to perform the accident analysis for the Pre-operational Safety Report (POSAR); (d) to provide a synthetic report addressing the validity range of codes models and correlations, the assessment against relevant tests results, the evidence of the user qualification, the modernisation and nodding scheme for the plant and the justification of used computer codes. Analyses results showed that all acceptance criteria were met with satisfactory margin and design of the NPP Mochovce is accurate. (author)

  1. Enhancement of seismic resistance of buildings

    Directory of Open Access Journals (Sweden)

    Claudiu-Sorin Dragomir

    2014-03-01

    Full Text Available The objectives of the paper are both seismic instrumentation for damage assessment and enhancing of seismic resistance of buildings. In according with seismic design codes in force the buildings are designed to resist at seismic actions. Due to the time evolution of these design provisions, there are buildings that were designed decades ago, under the less stringent provisions. The conceptual conformation is nowadays provided in all Codes of seismic design. According to the Code of seismic design P100-1:2006 the asymmetric structures do not have an appropriate seismic configuration; they have disadvantageous distribution of volumes, mass and stiffness. Using results of temporary seismic instrumentation the safety condition of the building may be assessed in different phases of work. Based on this method, the strengthening solutions may be identified and the need of seismic joints may be emphasised. All the aforementioned ideas are illustrated through a case study. Therefore it will be analysed the dynamic parameter evolution of an educational building obtained in different periods. Also, structural intervention scenarios to enhance seismic resistance will be presented.

  2. Evaluation of the atmospheric transport modeling tools used at the Ignalina NPP

    International Nuclear Information System (INIS)

    Jasiulionis, R.

    2004-01-01

    The atmospheric transport modeling system at the Ignalina NPP is part of the radionuclide monitoring system and is designed to help associate radionuclide network signals with possible source regions. The evaluation of the Atmospheric Transport Modeling Tools Used at the Ignalina NPP was based on results published in last years. The model based on semi-empirical formulae of turbulent diffusion in the air and local meteorological data were used for calculation of concentration of radionuclides emitted through the Ignalina NPP stack. Meteorological data (wind velocity and direction as well as temperature at a height of 2 and 30 meters) were obtained from the meteorological station of the Ignalina NPP at Visaginas. Using the Hybrid Single-Particle Lagrangian Integrated Trajectory (HYSPLIT) model, trajectories of air masses transport to the Ignalina NPP region (56.55 N, 26.57 E) during 95 hours at heights of 100, 500 and 1000 m were calculated. The dispersion in the ground-level air of emissions through the Ignalina NPP stack can also be calculated using the HYSPLIT model. The data on the radionuclide activity concentrations in the air at the measurements site and results of calculation can used for the establishment of their field in the Ignalina NPP region. (author)

  3. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, J.L.

    1998-01-01

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  4. Seismic design of RC buildings theory and practice

    CERN Document Server

    Manohar, Sharad

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  6. Perspectives for utilization of the experience from the electrical equipment modernization of Kozloduy NPP units 5 and 6 in case of construction of new NPP

    International Nuclear Information System (INIS)

    Bogdanov, D.

    2005-01-01

    In 2005 a governmental decision for continue the project for construction of Belene NPP was adopted. In case of start of this significant project in short terms, a good option will be to utilize the already achieved experience by the modernization program of units 5 and 6 of Kozloduy NPP. The process of construction of new NPP is associated with significant amount of engineering, civil construction works, equipment order and delivery, installation works, testing and commissioning activities. It is vital necessity to have optimal technology, strict organization for project implementation and excellent documentation. From technical point of view the experience from the modernization of units 5 and 6 of Kozloduy NPP will be very important in case of utilization of similar types of WWER 1000 units. According to the 'Feasibility study for construction of Belene NPP', the most favorable option is to accomplish the already partially constructed Unit 1 on the basis of upgraded WWER B-320 design - WWER 1000 B-392, in order to fulfill IEC 331-3A, IEC 332-3A, and of US NRC Regulatory Guide 1.70 and to construct the second unit on the basis of ASE WWER/1000 B-466. However, for any type of unit selected, the achieved experience will be an important basis. This article is focused mostly on the modernization aspects of some electrical systems of Kozloduy NPP

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

    International Nuclear Information System (INIS)

    Slagis, G.C.

    1996-01-01

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

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

    OpenAIRE

    Panasyuk Leonid; Kravchenko Galina; Trufanova Elena

    2017-01-01

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

  9. The NPP Isar comprehensive Aging Management Program

    International Nuclear Information System (INIS)

    Zander, Andre; Ertl, Stefan

    2012-01-01

    The majority of System, Structure and Components (SSC) in a nuclear power plants are designed to experience a service life, which is far above the intended design life. In most cases, only a small percentage of SSCs are subject to significant aging effects, which may affect the integrity or the function of the component. The process of aging management (AM) has the objective to monitor and control degradation effects which may compromise safety functions of the plant. And furthermore, to ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects. Safety-related aspects and the targeted high availability of the power plant as well as the requirements stipulated by German regulatory authorities prompted the operator of NPP ISAR to introduce an aging surveillance program. The NPP Isar as well as the German NPPs has to be following in the scope of aging management the KTA 1403 guideline. The NPP Isar surveillance program based on the KTA 1403 guideline covers the following aspects: - Scoping and screening of safety relevant Systems, Structures and Components (SSC); - Identification of possible degradation mechanisms for safety relevant SSC; - Ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects; - Transferability check of industry experience (internal and external events); - Annual preparation of an AM status report. (author)

  10. NPP Decommissioning: the concept; state of activities

    International Nuclear Information System (INIS)

    Nemytov, S.; Zimin, V.

    2001-01-01

    The main principles of NPP decommissioning concept in Russia are given. The conditions with fulfillment of works on NPP unit pre-decommissioning and decommissioning including: development of the normative documentation, creation of special fund for financing NPP decommissioning activities, deriving the Gosatomnadzor license for decommissioning of shut down NPP units, development of the equipment and technologies for waste and spent fuel management are presented. The decommissioning cost and labour intensity of one WWER-440 unit are shown. The practical works, executed on shut down units at Beloyarsk NPP (Unit1 and 2) and Novo Voronezh NPP (Unit 1 and 2) are outlined

  11. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2009-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  12. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2008-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  13. Results and development prospects for the Cernavoda NPP

    International Nuclear Information System (INIS)

    Rotaru, Ioan

    2001-01-01

    The new competitive market in the power sector offers favorable conditions for continual developing of Nuclear Power in Romania. The principal arguments in this respect are: the CANDU technology is a western technology recognized world wide both for its operation reliability and its economical performances. The results obtained by Cernavoda NPP Unit 1 over a period of four years and a half commercial operation confirms the validity of the CANDU project in Romania in the new conditions of competitive market. Accomplishing the Unit 2 of Cernavoda NPP represents for the time being one of the most cost-effective investment in the Romanian power sector. Also, the participation of the local industry to the project implementation is important as it ensures the initial heavy water and nuclear fuel charging, specific materials and equipment production, design and construction works, etc. Heaving in view all these favorable arguments, the construction of Unit 2 of Cernavoda NPP represents one of the highest priorities of the State Budget of the present administration in year 2001. The Nuclear Power Project in Romania, based on CANDU technology meets national as well as EU regulations concerning nuclear power and as such its development poses no significant problems both in the context of EU extension and in the frame of the developing competitive market in the power sector. For SNN SA the main objectives in view for the next stage are: operating Cernavoda NPP Unit 1 in conditions of highest nuclear safety and economical efficiency; finalizing the Cernavoda NPP Unit 2 and creation of an organisation for operating the two units; promoting of a functional system characteristic to the competitive market economy and open to privatization; creating opportunities for exporting electrical energy. The contribution contains the following sections: 1. General overlook; 2. Performance of the power production in the year 2000; 2.1. Unit 1 of Cernavoda NPP; 2.2. Nuclear Fuel Plant

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

    International Nuclear Information System (INIS)

    Nhorte Gomez, M.D.

    1994-01-01

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

  15. Analytical Study on the Beyond Design Seismic Capacity of Reinforced Concrete Shear Walls

    Energy Technology Data Exchange (ETDEWEB)

    Nugroho, Tino Sawaldi Adi [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chi, Ho-Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The OECD-NEA has organized an international benchmarking program to better understand this critical issue. The benchmark program provides test specimen geometry, test setup, material properties, loading conditions, recorded measures, and observations of the test specimens. The main objective of this research is to assess the beyond design seismic capacity of the reinforced concrete shear walls tested at the European Laboratory for Structural Assessment between 1997 and 1998 through participation in the OECD-NEA benchmark program. In this study, assessing the beyond design seismic capacity of reinforced concrete shear walls is performed analytically by comparing numerical results with experimental results. The seismic shear capacity of the reinforced concrete shear wall was predicted reasonably well using ABAQUS program. However, the proper calibration of the concrete material model was necessary for better prediction of the behavior of the reinforced concrete shear walls since the response was influenced significantly by the material constitutive model.

  16. Exploratory Shaft Seismic Design Basis Working Group report

    International Nuclear Information System (INIS)

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

    1990-08-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. NPP Mochovce units 1 and 2 diagnostic systems

    International Nuclear Information System (INIS)

    Heidenreich, S.

    1997-01-01

    In this paper the diagnostic systems (leak detection monitoring, vibration monitoring, lose parts monitoring, fatigue monitoring) of NPP Mochovce units 1 and 2 are presented. All of the designed diagnostic systems are personal computer based systems

  19. About choosing the power unit of NPP

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Mordashev, V.M.; Proshkin, A.A.; Stukalov, V.A.; Subbotin, S.A.; Tsibul'skij, V.F.; Chernilin, Yu.F.

    2008-01-01

    The present-day domestic nuclear power industry faces the necessity to make both process and economy related decisions. The mentioned decisions should cover all the aspects of the nuclear power complex. In particular, as of now, no substantiation of the present-day and the future requirements for the power unit of the WWER and fast reactor NPPs is available. The choice of the unit power of an NPP should involve consideration of all factors and should not be boiled down to the efforts to ensure the minimization of the investment and the operation constituents of the electric power generation costs. The aim of the present paper was to ensure elaboration of the recommendations to substantiate the optimal unit power of NPP power units based on the analysis of various factors to design, to construct and to operate an NPP (investment, time of construction, the unscheduled shutdown losses, the unification, the fabrication quality, the accident damage, etc.), and the electrical network potentialities and the regional electric power demands [ru

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

    International Nuclear Information System (INIS)

    Reddy, D.P.

    1983-04-01

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

  1. Calculation of behaviour of the Juragua NPP containment with code TRACOV/MOD1

    International Nuclear Information System (INIS)

    Castillo Alvarez, J.; Valle Cepero, R.; Luis, J.; San Roman, J.C.; Pomier, L.

    1996-01-01

    The containment of Juragua NPP has some unique features, which differ from the rest of the PWR reactors design. Those features impose additional requirements for its numerical simulation. In this paper is analyzed the behaviour of the Juragua NPP containment during accident situation with double ended break of the primary pipelines with flow in both direction using the code TRACOV/MOD1. The results are compared with obtained by the designer. The main restrictions of the code are identified

  2. Consideration on the applicability of the design seismic coefficient of a large cutting slope under the strong earthquake

    International Nuclear Information System (INIS)

    Ito, Hiroshi; Sawada, Yoshihiro; Satou, Kiyotaka

    1989-01-01

    In this study, the characteristic of equivalent seismic coefficient and the applicability of the design seismic coefficient of a large cutting rock slope around Nuclear Power Plant were examined by analytical parameter survey. As the results, the equivalent seismic coefficient by dynamic analysis become great with increase of transverse elastic wave velocity and the case of long period motion. That is, as the wave length of rock mass become longer, the equivalent seismic coefficient become great parabolically. Moreover, there is a inverse proportion relation between the ratio (dynamic safety factor/static safety factor) and wave length. In addition, the graph to forecast the dynamic sliding safety factor under the input seismic motion of the max. Acceleration 500 gal from the result of static simple method was proposed and the applicable range of design seismic coefficient of rock slope was indicated. (author)

  3. Performance-based seismic design of steel frames utilizing colliding bodies algorithm.

    Science.gov (United States)

    Veladi, H

    2014-01-01

    A pushover analysis method based on semirigid connection concept is developed and the colliding bodies optimization algorithm is employed to find optimum seismic design of frame structures. Two numerical examples from the literature are studied. The results of the new algorithm are compared to the conventional design methods to show the power or weakness of the algorithm.

  4. Full scale heavily reinforced concrete beam-column joints of NPP structures-quality assurance and construction in the laboratory

    International Nuclear Information System (INIS)

    Thandavamoorthy, T.S.; Vimalanandam, V.; Anandavalli, N.; Reddy, G.R.

    2003-01-01

    Under the current design philosophy, reactor structures are to be designed to withstand large inelastic deformation caused by strong and severe ground motion. The design of the main structural elements and their connections are to be such that they always fail in ductile mode. This will ensure large energy absorption capacity of the structures under seismic excitation and avoid sudden and brittle failure of the structure. Over the years, a number of experimental investigations have been carried out on RC beam- column joints to study their behaviour and strength. However, these studies mostly pertain to small scale joints of moment resisting frame of residential buildings and commercial complexes. Information on full scale joints existing in NPP structures are scanty. Therefore, experimental programme was planned in the laboratory to construct identical large sized joints with the same reinforcement percentage and detail as that of the existing joints in NPP structures built in 1960's. The cross-sectional dimensions of the beam and column of the joint were 610 mm x 915 mm. The beam was reinforced with 24 numbers of 36 mm bars. Column was reinforced with 4 numbers of 36 mm diameter bars and various other sizes. M25 grade concrete was used for casting of the specimen. The mix proportion was 1:1.657:2.63. The slump achieved was 75 mm to 100 mm. The health of the specimen was monitored by conducting ultrasonic testing with TICO instrument. The paper presents details of the size of the specimen and reinforcement, testing of steel bars for the evaluation of their mechanical properties, procedure of casting of specimen and its health monitoring. (author)

  5. Design of fuelling machine bridge and carriage to meet seismic qualification requirements

    International Nuclear Information System (INIS)

    Ghare, A.B.; Chhatre, A.G.; Vyas, A.K.; Bhambra, H.S.

    1996-01-01

    During each refuelling operation, the boundary of Primary heat transport system is extended up to Fuelling Machines. A breach in the pressure boundary of Fuelling Machine in this condition would cause a loss of coolant accident. Fuelling Machines are also used for transit storage of spent fuel bundles till discharged to fuel transfer system. Therefore, a fuelling machine, including its support structures, is required to be seismically qualified for both on-reactor ( coupled ) mode and off-reactor (uncoupled) mode. The fuelling machine carriage used in the first generation of Indian PHWRs is a mobile equipment on wheels moving over fixed rails. As this configuration was found unsuitable for withstanding strong seismic disturbances, a bridge type design with fixed columns was evolved for the next generation of reactors. Initially, the seismic analysis of the fuelling machine bridge and carriage was done using static structural analysis and values of natural frequencies for various structures were computed. The structures were suitably modified based on the results of this analysis. Subsequently, a detailed dynamic seismic analysis using finite element model has been completed for both coupled and uncoupled conditions. The qualification of the structure has been carried out as per ASME section 111 Division 1, sub section NF. Details of the significant design features, static and dynamic analysis, results and conclusions are given in the presentation. (author). 4 refs., 4 tabs., 7 figs

  6. Design of fuelling machine bridge and carriage to meet seismic qualification requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ghare, A B; Chhatre, A G; Vyas, A K; Bhambra, H S [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    During each refuelling operation, the boundary of Primary heat transport system is extended up to Fuelling Machines. A breach in the pressure boundary of Fuelling Machine in this condition would cause a loss of coolant accident. Fuelling Machines are also used for transit storage of spent fuel bundles till discharged to fuel transfer system. Therefore, a fuelling machine, including its support structures, is required to be seismically qualified for both on-reactor ( coupled ) mode and off-reactor (uncoupled) mode. The fuelling machine carriage used in the first generation of Indian PHWRs is a mobile equipment on wheels moving over fixed rails. As this configuration was found unsuitable for withstanding strong seismic disturbances, a bridge type design with fixed columns was evolved for the next generation of reactors. Initially, the seismic analysis of the fuelling machine bridge and carriage was done using static structural analysis and values of natural frequencies for various structures were computed. The structures were suitably modified based on the results of this analysis. Subsequently, a detailed dynamic seismic analysis using finite element model has been completed for both coupled and uncoupled conditions. The qualification of the structure has been carried out as per ASME section 111 Division 1, sub section NF. Details of the significant design features, static and dynamic analysis, results and conclusions are given in the presentation. (author). 4 refs., 4 tabs., 7 figs.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. Feed water distribution pipe replacement at Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S.; Elsing, B. [Imatran Voima Loviisa NPP (Finland)

    1995-12-31

    Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

  9. Feed water distribution pipe replacement at Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S; Elsing, B [Imatran Voima Loviisa NPP (Finland)

    1996-12-31

    Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

  10. Use of the deterministic safety analyses in support to the NPP Krsko modification

    International Nuclear Information System (INIS)

    Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

    2004-01-01

    The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(author)

  11. Application of Nonlinear Seismic Soil-Structure Interaction Analysis for Identification of Seismic Margins at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Varma, Amit H.; Seo, Jungil; Coleman, Justin Leigh

    2015-01-01

    Seismic probabilistic risk assessment (SPRA) methods and approaches at nuclear power plants (NPP) were first developed in the 1970s and aspects of them have matured over time as they were applied and incrementally improved. SPRA provides information on risk and risk insights and allows for some accounting for uncertainty and variability. As a result, SPRA is now used as an important basis for risk-informed decision making for both new and operating NPPs in the US and in an increasing number of countries globally. 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 some instances the current SPRA approach contains large uncertainties, 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). INL has an advanced SPRA research and development (R&D) activity that will identify areas in the calculation process that contain significant uncertainties. One current area of focus is the use of nonlinear soil-structure interaction (NLSSI) analysis methods to accurately capture: 1) nonlinear soil behavior and 2) gapping and sliding between the NPP and soil. The goal of this study is to compare numerical NLSSI analysis results with recorded earthquake ground motions at Fukushima Daichii (Great Tohuku Earthquake) and evaluate the sources of nonlinearity contributing to the observed reduction in peak acceleration. Comparisons are made using recorded data in the free-field (soil column with no structural influence) and recorded data on the NPP basemat (in-structure response). Results presented in this study should identify areas of focus for future R&D activities with the goal of minimizing uncertainty in SPRA calculations. This is not a validation activity since there are too many sources of uncertainty that a numerical analysis would need

  12. Opinion of the Bulgarian Academy of Sciences on the Belene NPP construction

    International Nuclear Information System (INIS)

    1990-01-01

    The chapter states in a concise form the essence of the independent teams study results on the project of construction of the new Belene nuclear power plant in Bulgaria on the Danube river. The general opinion is negative. The studies on the energy-economy interactions show that under the new conditions of social, economic and technological development, the energy balance of the country is achievable up to 2000 without the Belene NPP. With the construction of the Belene NPP the energy development for the next 15 years would be more expensive by 45 million leva total costs and by 22-26 million leva anti-import production than in the case of the studied power supply variants for less-energy-consuming development. The technical project study reveals the insufficient substantiation of different technological safety aspects of the plant. No significant negative consequences related to the radiation safety in case of normal operation are foreseen but as regards the possible accidents with off-site risks, the site is not well chosen. The waste storage is satisfactorily designed. There is the pending problem of radioactive waste disposal for the country as a whole. The economic efficiency study presents a payback period longer than the service life of the plant and, taking into account the decommissioning costs, the analyses are unfavorable. The chosen site does not have any resource of seismic safety. The project does not provide a clear picture of the environmental impacts. Without an ecological study as a part of the project the construction is not substantiated. According to the public opinion pools the construction is unacceptable. (R.Ts.)

  13. The Service Life Evaluation for Concrete Structure of NPP

    International Nuclear Information System (INIS)

    Lee, Choon Min; Kim, Seong Soo; Bae, Sung Hwan; Sik, Yoon Eui

    2014-01-01

    Prolonged exposure to the marine environment degrades the durability of concrete and shortens the service life of concrete due to degradation factors such as chloride, carbonation, freezing and thawing, sulfate. Therefore, many country's organizations like the Korea Concrete Institute (KCI), the American Concrete Institute (ACI), the International Federation for Structural Concrete (FIB), the American Society for Testing and Material (ASTM) which recognized the seriousness of this deterioration proposed equation models to evaluate the service life for the concrete structures. As a result, this paper is to especially consider the service life evaluation using these equation models for concrete structure of NPP in Korea compared with 60 years as a design service life. The concrete durability evaluation for Shin-Hanul NPP units 1 and 2 is carried out by using typically proposed assessment models in domestic and foreign standard. It is confirmed that the service life of concrete durability for each of deterioration factors is significantly higher than 60 years as a design service life. As a result, the study of combined deterioration for the concrete structures of NPP in future is positively necessary and maintenance control through regular monitoring should be conducted to secure safety margin basis

  14. Seismic design of a uranium conversion plant building

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. The contract for the completion and commissioning of Cernavoda NPP - Unit

    International Nuclear Information System (INIS)

    2001-01-01

    On May 18, 2001 the commercial contract for the completion and commissioning of the Cernavoda NPP Unit 2 has been signed by Nuclearelectrica National Company (SNN SA) and its partners, AECL-Canada and ANSALDO-Italy. This contract represents a continuation of the cooperation with AECL-Canada (the owner of the licence and the designer of the nuclear part of Cernavoda NPP Unit1) and ANSALDO-Italy (the designer of the conventional part of Cernavoda NPP Unit1). The completion is scheduled to last for 54 months (4.5 years) since the date the contract takes effect and the nuclear reactor start-up (the first criticality) is scheduled for the 49 th month. The total negotiated cost for the completion was estimated to amount to 689 million dollars, against the 750 million dollars initially estimated. The works at the site will be managed by an integrated staff team of specialized personnel form AECL-Canada (111 positions), ANSALDO-Italy (72 positions) and Romania (1200 positions)

  16. NPP service life management

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    Problems of NPP service life management and service life prolongation are reviewed. Methods for the prolongation of the French NPP service life are discussed, priority directions of nuclear block service life management in regard to aging in the context of the European program of investigation into the materials aging are identified. Questions of the provision of the 60 years service life of the Mihama 1 block (Japan) and decision of the problem of the control equipment aging in Great Britain are discussed. Situation with the prolongation of licenses on the NPP operation in the USA and Spain is considered [ru

  17. 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

  18. 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

  19. 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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  1. Improved Simplified Methods for Effective Seismic Analysis and Design of Isolated and Damped Bridges in Western and Eastern North America

    Science.gov (United States)

    Koval, Viacheslav

    The seismic design provisions of the CSA-S6 Canadian Highway Bridge Design Code and the AASHTO LRFD Seismic Bridge Design Specifications have been developed primarily based on historical earthquake events that have occurred along the west coast of North America. For the design of seismic isolation systems, these codes include simplified analysis and design methods. The appropriateness and range of application of these methods are investigated through extensive parametric nonlinear time history analyses in this thesis. It was found that there is a need to adjust existing design guidelines to better capture the expected nonlinear response of isolated bridges. For isolated bridges located in eastern North America, new damping coefficients are proposed. The applicability limits of the code-based simplified methods have been redefined to ensure that the modified method will lead to conservative results and that a wider range of seismically isolated bridges can be covered by this method. The possibility of further improving current simplified code methods was also examined. By transforming the quantity of allocated energy into a displacement contribution, an idealized analytical solution is proposed as a new simplified design method. This method realistically reflects the effects of ground-motion and system design parameters, including the effects of a drifted oscillation center. The proposed method is therefore more appropriate than current existing simplified methods and can be applicable to isolation systems exhibiting a wider range of properties. A multi-level-hazard performance matrix has been adopted by different seismic provisions worldwide and will be incorporated into the new edition of the Canadian CSA-S6-14 Bridge Design code. However, the combined effect and optimal use of isolation and supplemental damping devices in bridges have not been fully exploited yet to achieve enhanced performance under different levels of seismic hazard. A novel Dual-Level Seismic

  2. Seismic isolation systems designed with distinct multiple frequencies

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  3. 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)

  4. Development of NPP Monitoring and Operation Support Technology

    International Nuclear Information System (INIS)

    Lee, Jung Woon; Park, Jae Chang; Lee, Yong Hee

    2008-04-01

    During the first stage (2001.7.1-2004.6.30), we developed general human factors design guidelines VDU-based workstations, PMAS alarm display design guidelines, PMAS SPADES display design guidelines, and the revision of KHNP HFE guidelines (HF-010), which have been applied to domestic NPP designs. We also supported other KNICS projects by performing RPS COM design reviews, development of RPS COM Style Guide, and a review of CEDMCS cabinet operator module display design. We developed the ADIOS prototype, NPP performance analysis systems for YGN No.1, 2 plants and Kori No. 2 plant, alarm cause tracking systems for Kori No. 2 plant and OPR1000, and signal fault detection and diagnosis methods for deaerators and steam generators. During the second stage(2004.7.1-2008.4.30), we supported other KNICS projects by reviewing RPS COM display designs three times, developing ESF-CCS COM style guides and reviewing ESF-CCS COM display design, reviewing CRCS LOM and PCS MTP display designs, and developing requirements for DCS GUI components. We also developed integrated style guide for I and C cabinet operator module display designs. In cooperative research with KOPEC-AE, we developed basic technologies for advanced HSI design including task analysis methods, an information and control requirements database, display design criteria, a HSI prototype with its evaluation, and methods for human factors engineering verification and validation

  5. Development of NPP Monitoring and Operation Support Technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Woon; Park, Jae Chang; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2008-04-15

    During the first stage (2001.7.1-2004.6.30), we developed general human factors design guidelines VDU-based workstations, PMAS alarm display design guidelines, PMAS SPADES display design guidelines, and the revision of KHNP HFE guidelines (HF-010), which have been applied to domestic NPP designs. We also supported other KNICS projects by performing RPS COM design reviews, development of RPS COM Style Guide, and a review of CEDMCS cabinet operator module display design. We developed the ADIOS prototype, NPP performance analysis systems for YGN No.1, 2 plants and Kori No. 2 plant, alarm cause tracking systems for Kori No. 2 plant and OPR1000, and signal fault detection and diagnosis methods for deaerators and steam generators. During the second stage(2004.7.1-2008.4.30), we supported other KNICS projects by reviewing RPS COM display designs three times, developing ESF-CCS COM style guides and reviewing ESF-CCS COM display design, reviewing CRCS LOM and PCS MTP display designs, and developing requirements for DCS GUI components. We also developed integrated style guide for I and C cabinet operator module display designs. In cooperative research with KOPEC-AE, we developed basic technologies for advanced HSI design including task analysis methods, an information and control requirements database, display design criteria, a HSI prototype with its evaluation, and methods for human factors engineering verification and validation.

  6. A comparison of the Kozloduy NPP and Temelin NPP I and C projects

    International Nuclear Information System (INIS)

    Cook, B.M.

    2004-01-01

    Because Kozloduy NPP and Temelin NPP are both VVER 1000 plants of roughly the same vintage, they are very similar in design. However, from the viewpoint of their I and C modernization projects, there are significant differences between these plants. Some of these differences stem from the evolution of I and C technology over the relatively short period between the two projects. Other differences arise from the fact that the Kozloduy project is a phased upgrade of the I and C systems in an operating plant while the Temelin project was a 'one time' installation of the entire plant I and C system. This paper discusses these differences as well as trends in the nuclear I and C field that will shape the industry in the future. In addition to technology evolution, the comparative advantages or problems in phased upgrade versus 'one time' installations are discussed. Conclusions drawn provide insight for the planning of future I and C upgrades in VVERs and other types of nuclear power plants. (author)

  7. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  8. A performance goal-based seismic design philosophy for waste repository facilities

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1994-02-01

    A performance goal-based seismic design philosophy, compatible with DOE's present natural phenomena hazards mitigation and ''graded approach'' philosophy, has been proposed for high level nuclear waste repository facilities. The rationale, evolution, and the desirable features of this method have been described. Why and how the method should and can be applied to the design of a repository facility are also discussed

  9. Calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled WWER-type reactor NPPs. Final report. From 01 May 1997 - 30 April 1998

    International Nuclear Information System (INIS)

    Kaznovski, S.

    1999-01-01

    In conformity with request from Paks NPP in July 1997 the additional dynamic tests of the following units of equipment were carried out: Heat exchanger (Diesel Generator Station); Jalousie (Diesel Generator Station); Axial ventilator (Main Building, gallery of electric equipment). The natural frequencies of equipment were obtained as results of these testing. For all tested units the calculations of seismic resistance were carried out. The method of finite elements and special computer programmes of KBGU (ISPA, and others) were used. Calculations were carried out in accordance with modem standards. The calculations of tested equipment seismic resistance are carried out based on the obtained experimental data and NPP floor response spectra. The necessary recommendations for unsatisfactory equipment are worked out

  10. Operator reliability analysis during NPP small break LOCA

    International Nuclear Information System (INIS)

    Zhang Jiong; Chen Shenglin

    1990-01-01

    To assess the human factor characteristic of a NPP main control room (MCR) design, the MCR operator reliability during a small break LOCA is analyzed, and some approaches for improving the MCR operator reliability are proposed based on the analyzing results

  11. The landscape basis of the complex monitoring of 30-km zone of NPP

    International Nuclear Information System (INIS)

    Malisheva, L.L.; Romanchuk, S.P.; Schur, U.V.; Rybalko, S.I.; Proskura, N.I.; Lury, D.I.

    1992-01-01

    This publication reviews works on the problem of creation the landscape basis of complex monitoring in 30 km zone of NPP(nuclear power plants). The objectives, subject and designation of monitoring system are considered. The monitoring network and organization of systematical observations are justified. The structure of radiation monitoring in NPP 30 km zone is considered in details. (author)

  12. Search and Retrieval of Foreign Objects for the Steam Generator of Wolsung NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Kyung-Ho [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    We developed a foreign object search and retrieval (FOSAR) system for Wolsung NPP unit 1 steam generators. The steam generators of Wolsung NPP unit 1 have one 2.5 inch hand hole and two 4 inch hand holes. The FOSAR system was designed to be installed through 4 inch hand holes. Using permanent magnet, the FOSAR system was firmly attached to the vertical annulus wall of the steam generator. We successfully developed the FOSAR system for Wolsung NPP unit 1. Using the developed FOSAR system, technicians successfully found and removed various foreign objects. Most of the foreign objects, we found, were made of carbon steel sheet, therefore magnet tool was the most useful to remove it. Alligator tool was sometimes used. Based on the experience during the FOSAR activities, we are developing a lancing system for Wolsung NPP unit 1. It will be designed and manufactured until November 2016.

  13. Search and Retrieval of Foreign Objects for the Steam Generator of Wolsung NPP Unit 1

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Kyung-Ho

    2016-01-01

    We developed a foreign object search and retrieval (FOSAR) system for Wolsung NPP unit 1 steam generators. The steam generators of Wolsung NPP unit 1 have one 2.5 inch hand hole and two 4 inch hand holes. The FOSAR system was designed to be installed through 4 inch hand holes. Using permanent magnet, the FOSAR system was firmly attached to the vertical annulus wall of the steam generator. We successfully developed the FOSAR system for Wolsung NPP unit 1. Using the developed FOSAR system, technicians successfully found and removed various foreign objects. Most of the foreign objects, we found, were made of carbon steel sheet, therefore magnet tool was the most useful to remove it. Alligator tool was sometimes used. Based on the experience during the FOSAR activities, we are developing a lancing system for Wolsung NPP unit 1. It will be designed and manufactured until November 2016

  14. Effect of URM infills on seismic vulnerability of Indian code designed RC frame buildings

    Science.gov (United States)

    Haldar, Putul; Singh, Yogendra; Paul, D. K.

    2012-03-01

    Unreinforced Masonry (URM) is the most common partitioning material in framed buildings in India and many other countries. Although it is well-known that under lateral loading the behavior and modes of failure of the frame buildings change significantly due to infill-frame interaction, the general design practice is to treat infills as nonstructural elements and their stiffness, strength and interaction with the frame is often ignored, primarily because of difficulties in simulation and lack of modeling guidelines in design codes. The Indian Standard, like many other national codes, does not provide explicit insight into the anticipated performance and associated vulnerability of infilled frames. This paper presents an analytical study on the seismic performance and fragility analysis of Indian code-designed RC frame buildings with and without URM infills. Infills are modeled as diagonal struts as per ASCE 41 guidelines and various modes of failure are considered. HAZUS methodology along with nonlinear static analysis is used to compare the seismic vulnerability of bare and infilled frames. The comparative study suggests that URM infills result in a significant increase in the seismic vulnerability of RC frames and their effect needs to be properly incorporated in design codes.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  16. 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

  17. Towards safe and economic seismic design of cooling towers of extreme height

    International Nuclear Information System (INIS)

    Kraetzig, W.B.; Meskouris, K.

    1979-01-01

    Nuclear power plants are being increasingly equipped with natural draught cooling towers of heights greater than 160 m. In many arid zones, where high natural draught cooling towers with dry cooling systems are being projected, wind loads are relativelly small while site seismicity is relatively high. Thus the ability of the tower to withstand earthquake induced forces governs its design. On the other hand, most reinforced concrete cooling towers of extreme height built so far were designed to withstand high wind loads and moderate earthquake loads. The effects of special structural measures for obtaining an economic design, such as the introduction of ring stiffened shells, have been studied mainly for those towers. In view of the previous aspects it is the purpose of this paper to analyze the effects of various structural measures and other parameters on the seismic response of such high cooling towers. (orig.)

  18. Safety parameter display system for Kalinin NPP

    International Nuclear Information System (INIS)

    Andreev, V.I.; Videneev, E.N.; Tissot, J.C.; Joonekindt, D.; Davidenko, N.N.; Shaftan, G.I.; Dounaev, V.G.; Neboyan, V.T.

    1995-01-01

    The paper discusses the safety parameter display system (SPDS), which is being designed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. As compared with the traditional scope of functions of this kind of systems, the functionality of KlnNPP SPDS is significantly expanded due to the inclusion in it the operator support function ''computerized procedures''. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis center and to the crisis center of the State utility organization concern ''Rosenergoatom''. (author). 3 refs, 6 figs, 1 tab

  19. Modernization programme at Dukovany NPP

    International Nuclear Information System (INIS)

    Trnka, M.

    2000-01-01

    The main goal of each NPP is to produce electricity safely, economically and without influence to environment. For Dukovany NPP it means to upgrade all documentation and perform the Equipment Upgrading Programme. All these activities are time and money consuming and therefore the determination of priority of all items was necessary. In the presentation there are mentioned some important changes in documentation, results of PSA studies and reason for Equipment Upgrading Programme performance. It was selected the most important item from the list of Equipment Upgrading Programme the I and C upgrading. Management has decided that Dukovany NPP will become among the best NPPs with WWER type of reactor. It seems this decision is the best way how to extend lifetime of the NPP. (author)

  20. NRC systematic evaluation program: seismic review

    International Nuclear Information System (INIS)

    Levin, H.A.

    1980-01-01

    The NRC Systematic Evaluation Program is currently making an assessment of the seismic design safety of 11 older nuclear power plant facilities. The general review philosophy and review criteria relative to seismic input, structural response, and equipment functionability are presented, including the rationale for the development of these guidelines considering the significant evolution of seismic design criteria since these plants were originally licensed. Technical approaches thought more realistic in light of current knowledge are utilized. Initial findings for plants designed to early seismic design procedures suggest that with minor exceptions, these plants possess adequate seismic design margins when evaluated against the intent of current criteria. However, seismic qualification of electrical equipment has been identified as a subject which requires more in-depth evaluation