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

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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

Seismic evaluation of commercial plutonium fabrication plants in the United States

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Seismic capacities of masonry walls at the big rock point nuclear generating plant

International Nuclear Information System (INIS)

Wesley, D.A.; Bunon, H.; Jenkins, R.B.

1984-01-01

An evaluation to determine the ability of selected concrete block walls in the vicinity of essential equipment to withstand seismic excitation was conducted. The seismic input to the walls was developed in accordance with the Systematic Evaluation Program (SEP) site-specific response spectra for the site. Time-history inputs to the walls were determined from the response of the turbine building complex. Analyses were performed to determine the capacities of the walls to withstand both in-plane and transverse seismic loads. Transverse load capacities were determined from time-history analyses of nonlinear two-dimensional analytical models of the walls. Separate inputs were used at the tops and bottoms of the walls to reflect the amplification through the building. The walls were unreinforced vertically with one exception, and have unsupported heights as high as 20'-8''. Also, cantilever walls as high as 11'-2'' were included in the evaluation. Factors of safety based on stability of the walls were determined for the transverse response, and on code allowable stresses (Reference 1) for the in-plane response

Aging evaluation of class 1E batteries: Seismic testing

International Nuclear Information System (INIS)

Edson, J.L.

1990-08-01

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

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Status report on seismic re-evaluation

International Nuclear Information System (INIS)

1998-01-01

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

Seismic Capacity Estimation of Steel Piping Elbow under Low-cycle Fatigue Loading

Energy Technology Data Exchange (ETDEWEB)

Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of); Hahm, Dae Gi [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In some cases, this large relative displacement can increase seismic risk of the isolated facility. Especially, a inelastic behavior of crossover piping system to connect base isolated building and fixed base building can caused by a large relative displacement. Therefore, seismic capacity estimation for isolated piping system is needed to increase safety of nuclear power plant under seismic condition. Dynamic behavior analysis of piping system under seismic condition using shake table tests was performed by Touboul et al in 1995. In accordance with their study, plastic behavior could be occurred at pipe elbow under seismic condition. Experimental researches for dynamic behavior of typical piping system in nuclear power plant have been performed for several years by JNES(Japan Nuclear Energy Safety Organization) and NUPEC(Nuclear Power Engineering Corporation). A low cycle ratcheting fatigue test was performed with scaled model of elbow which is a weakest component in piping system by Mizuno et al. In-plane cyclic loading tests under internal pressure condition were performed to evaluate the seismic capacity of the steel piping elbow. Leakage phenomenon occurred on and near the crown in piping elbow. Those cracks grew up in axial direction. The fatigue curve was estimated from test results. In the fatigue curve, loading amplitude exponentially decreased as the number of cycles increased. A FEM model of piping elbow was modified with test results. The relationships between displacement and force from tests and numerical analysis was well matched.

Evaluation of seismic margins for an in-plant piping system

International Nuclear Information System (INIS)

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

1991-01-01

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

Destructiveness criteria for seismic risk evaluation of nuclear power plant

International Nuclear Information System (INIS)

Saragoni, G.R.

1995-01-01

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

A proposal for seismic evaluation index of mid-rise existing RC buildings in Afghanistan

Science.gov (United States)

Naqi, Ahmad; Saito, Taiki

2017-10-01

Mid-rise RC buildings gradually rise in Kabul and entire Afghanistan since 2001 due to rapid increase of population. To protect the safety of resident, Afghan Structure Code was issued in 2012. But the building constructed before 2012 failed to conform the code requirements. In Japan, new sets of rules and law for seismic design of buildings had been issued in 1981 and severe earthquake damage was disclosed for the buildings designed before 1981. Hence, the Standard for Seismic Evaluation of RC Building published in 1977 has been widely used in Japan to evaluate the seismic capacity of existing buildings designed before 1981. Currently similar problem existed in Afghanistan, therefore, this research examined the seismic capacity of six RC buildings which were built before 2012 in Kabul by applying the seismic screening procedure presented by Japanese standard. Among three screening procedures with different capability, the less detailed screening procedure, the first level of screening, is applied. The study founds an average seismic index (IS-average=0.21) of target buildings. Then, the results were compared with those of more accurate seismic evaluation procedures of Capacity Spectrum Method (CSM) and Time History Analysis (THA). The results for CSM and THA show poor seismic performance of target buildings not able to satisfy the safety design limit (1/100) of the maximum story drift. The target buildings are then improved by installing RC shear walls. The seismic indices of these retrofitted buildings were recalculated and the maximum story drifts were analyzed by CSM and THA. The seismic indices and CSM and THA results are compared and found that building with seismic index larger than (IS-average =0.4) are able to satisfy the safety design limit. Finally, to screen and minimize the earthquake damage over the existing buildings, the judgement seismic index (IS-Judgment=0.5) for the first level of screening is proposed.

Preliminary Seismic Performance Evaluation of RPS Cabinet in a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

Effect of the selected seismic energy dissipation capacity on the materials quantity for reinforced concrete walls

Directory of Open Access Journals (Sweden)

José Miguel Benjumea Royero

2017-02-01

Full Text Available Context: Regarding their design of reinforced concrete structural walls, the Colombian seismic design building code allows the engineer to select one of the three seismic energy dissipation capacity (ordinary, moderate, and special depending on the seismic hazard of the site. Despite this, it is a common practice to choose the minor requirement for the site because it is thought that selecting a higher requirement will lead to larger structural materials amounts and, therefore, cost increments.  Method: In this work, an analytical study was performed in order to determine the effect of the selected energy dissipation capacity on the quantity of materials and ductility displacement capacity of R/C walls. The study was done for a region with low seismic hazard, mainly because this permitted to explore and compare the use of the three seismic energy dissipations capacities. The effect of different parameters such as the wall total height and thickness, the tributary loaded area, and the minimum volumetric steel ratio were studied. Results: The total amount of steel required for the walls with moderate and special energy dissipation capacity corresponds, on average, to 77% and 89%, respectively, of the quantity required for walls with minimum capacity. Conclusions: it is possible to achieve reductions in the total steel required weight when adopting either moderated or special seismic energy dissipation instead of the minimum capacity.  Additionally, a significant increment in the seismic ductility displacements capacity of the wall was obtained.

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

International Nuclear Information System (INIS)

Moon, Ki-Hoon; Han, Sang-Whan

2008-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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.

Seismic evaluation of reinforced masonry walls

International Nuclear Information System (INIS)

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

1984-01-01

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

Estimation of Cyclic Interstory Drift Capacity of Steel Framed Structures and Future Applications for Seismic Design

Directory of Open Access Journals (Sweden)

Edén Bojórquez

2014-01-01

Full Text Available Several studies have been devoted to calibrate damage indices for steel and reinforced concrete members with the purpose of overcoming some of the shortcomings of the parameters currently used during seismic design. Nevertheless, there is a challenge to study and calibrate the use of such indices for the practical structural evaluation of complex structures. In this paper, an energy-based damage model for multidegree-of-freedom (MDOF steel framed structures that accounts explicitly for the effects of cumulative plastic deformation demands is used to estimate the cyclic drift capacity of steel structures. To achieve this, seismic hazard curves are used to discuss the limitations of the maximum interstory drift demand as a performance parameter to achieve adequate damage control. Then the concept of cyclic drift capacity, which incorporates information of the influence of cumulative plastic deformation demands, is introduced as an alternative for future applications of seismic design of structures subjected to long duration ground motions.

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

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.

Seismic assessment of air-cooled type emergency electric power supply system

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

JNES initiated seismic assessment programs to develop seismic review criterions for the air-cooled system (diesel generator, gas turbine generator), which will be newly installed for enhancing the diversity of emergency electric power supply system. Five principal subjects are involved in the programs: two subjects for fiscal 2011 and three ones for fiscal 2012 and 2013. The summary of outcomes is as follows: 1) Past capacity test data and related technical issues (2011). Seismic capacity data obtained from past seismic shaking tests were investigated. 2) Test programs based on the investigation of system specification (2011). Design specifications for the air-cooled system were investigated. 3) Large Air Fin Cooler (AFC) one unit model seismic capacity test and quantitative seismic capacity evaluation. AFC one unit model seismic capacity tests were conducted and quantitative seismic capacities were evaluated. (author)

Seismic assessment of air-cooled type emergency electric power supply system

International Nuclear Information System (INIS)

2013-01-01

JNES initiated seismic assessment programs to develop seismic review criterions for the air-cooled system (diesel generator, gas turbine generator), which will be newly installed for enhancing the diversity of emergency electric power supply system. Five principal subjects are involved in the programs: two subjects for fiscal 2011 and three ones for fiscal 2012 and 2013. The summary of outcomes is as follows: 1) Past capacity test data and related technical issues (2011). Seismic capacity data obtained from past seismic shaking tests were investigated. 2) Test programs based on the investigation of system specification (2011). Design specifications for the air-cooled system were investigated. 3) Large Air Fin Cooler (AFC) one unit model seismic capacity test and quantitative seismic capacity evaluation. AFC one unit model seismic capacity tests were conducted and quantitative seismic capacities were evaluated. (author)

Seismic evaluation and strengthening of Bohunice nuclear power plant structures

International Nuclear Information System (INIS)

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

2001-01-01

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

Consideration of time-evolving capacity distributions and improved degradation models for seismic fragility assessment of aging highway bridges

International Nuclear Information System (INIS)

Ghosh, Jayadipta; Sood, Piyush

2016-01-01

This paper presents a methodology to develop seismic fragility curves for deteriorating highway bridges by uniquely accounting for realistic pitting corrosion deterioration and time-dependent capacity distributions for reinforced concrete columns under chloride attacks. The proposed framework offers distinct improvements over state-of-the-art procedures for fragility assessment of degrading bridges which typically assume simplified uniform corrosion deterioration model and pristine limit state capacities. Depending on the time in service life and deterioration mechanism, this study finds that capacity limit states for deteriorating bridge columns follow either lognormal distribution or generalized extreme value distributions (particularly for pitting corrosion). Impact of column degradation mechanism on seismic response and fragility of bridge components and system is assessed using nonlinear time history analysis of three-dimensional finite element bridge models reflecting the uncertainties across structural modeling parameters, deterioration parameters and ground motion. Comparisons are drawn between the proposed methodology and traditional approaches to develop aging bridge fragility curves. Results indicate considerable underestimations of system level fragility across different damage states using the traditional approach compared to the proposed realistic pitting model for chloride induced corrosion. Time-dependent predictive functions are provided to interpolate logistic regression coefficients for continuous seismic reliability evaluation along the service life with reasonable accuracy. - Highlights: • Realistic modeling of chloride induced corrosion deterioration in the form of pitting. • Time-evolving capacity distribution for aging bridge columns under chloride attacks. • Time-dependent seismic fragility estimation of highway bridges at component and system level. • Mathematical functions for continuous tracking of seismic fragility along service

Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

International Nuclear Information System (INIS)

Stevenson, J.D.

1995-01-01

The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, components for which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation

Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Stevenson, J D [Stevenson and Associates, Cleveland, OH (United States)

1995-07-01

The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, componentsfor which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation.

Seismic evaluation of a hot cell structure

International Nuclear Information System (INIS)

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

1995-01-01

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

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

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)

Nonlinear seismic analysis of a thick-walled concrete canyon structure

International Nuclear Information System (INIS)

Winkel, B.V.; Wagenblast, G.R.

1989-01-01

Conventional linear seismic analyses of a thick-walled lightly reinforced concrete structure were found to grossly underestimate its seismic capacity. Reasonable estimates of the seismic capacity were obtained by performing approximate nonlinear spectrum analyses along with static collapse evaluations. A nonlinear time history analyses is planned as the final verification of seismic adequacy

Integrated system for seismic evaluations

International Nuclear Information System (INIS)

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

1989-01-01

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

Development of seismic PSA methodology at JAERI

International Nuclear Information System (INIS)

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

1995-01-01

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

Assessment of seismic margin calculation methods

International Nuclear Information System (INIS)

Kennedy, R.P.; Murray, R.C.; Ravindra, M.K.; Reed, J.W.; Stevenson, J.D.

1989-03-01

Seismic margin review of nuclear power plants requires that the High Confidence of Low Probability of Failure (HCLPF) capacity be calculated for certain components. The candidate methods for calculating the HCLPF capacity as recommended by the Expert Panel on Quantification of Seismic Margins are the Conservative Deterministic Failure Margin (CDFM) method and the Fragility Analysis (FA) method. The present study evaluated these two methods using some representative components in order to provide further guidance in conducting seismic margin reviews. It is concluded that either of the two methods could be used for calculating HCLPF capacities. 21 refs., 9 figs., 6 tabs

Upgrading of seismic capacity of WWER equipment including reactor control rods and distribution systems

International Nuclear Information System (INIS)

Kostarev, V.

1993-01-01

This paper concerns the experience of CKTI VIBROSEISM firm (CVS) in seismic capacity upgrading of the WWER equipment and pipings. During the last 15 years CVS has accumulated a lot of experimental and analysis results on many of WWER units 'in former Soviet Union. That might be very useful in seismic protecting of East European Nuclear Power Plants. (author)

Seismic margins review of nuclear power plants: Fragility aspects

International Nuclear Information System (INIS)

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

1987-01-01

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

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

Systems considerations in seismic margin evaluations

International Nuclear Information System (INIS)

Buttermer, D.R.

1987-01-01

Increasing knowledge in the geoscience field has led to the understanding that, although highly unlikely, it is possible for a nuclear power plant to be subjected to earthquake ground motion greater than that for which the plant was designed. While it is recognized that there are conservatisms inherent in current design practices, interest has developed in evaluating the seismic risk of operating plants. Several plant-specific seismic probabilistic risk assessments (SPRA) have been completed to address questions related to the seismic risk of a plant. The results from such SPRAs are quite informative, but such studies may entail a considerable amount of expensive analysis of large portions of the plant. As an alternative to an SPRA, it may be more practical to select an earthquake level above the design basis for which plant survivability is to be demonstrated. The principal question to be addressed in a seismic margin evaluation is: At what ground motion levels does one have a high confidence that the probability of seismically induced core damage is sufficiently low? In a seismic margin evaluation, an earthquake level is selected (based on site-specific geoscience considerations) for which a stable, long-term safe shutdown condition is to be demonstrated. This prespecified earthquake level is commonly referred to as the seismic margin earthquake (SME). The Electric Power Research Institute is currently supporting a research project to develop procedures for use by the utilities to allow them to perform nuclear plant seismic margin evaluations. This paper describes the systems-related aspects of these procedures

Seismic capacities of existing nuclear power plant structures

International Nuclear Information System (INIS)

Wesley, D.A.; Hashimoto, P.S.; Narver, R.B.

1983-01-01

The paper presents a discussion of the more important conservatisms and some of the results obtained when this methodology has been applied to various nuclear plants. Results are shown for both BWR and PWR plants, on both rock and soil sites, and for plants and soil sites, and for plants that were designed in the late 1960s to plants that have yet to load fuel. Safe shutdown earthquake design levels of 0.1 g to 0.25 g were used for these plants. Overall median structural factors of safety for the lowest significant seismic failure capacity at each plant ranged from 3.5 to 8.5. The lowest containment-related failure capacity at each plant ranged from 4.6 to 31. The types of failure corresponding to each safety factor are also tabulated. (orig./HP)

Integrated system for seismic evaluations

International Nuclear Information System (INIS)

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

1989-01-01

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

Seismic bearing capacity of strip footings on rock masses using the Hoek–Brown failure criterion

Directory of Open Access Journals (Sweden)

Amin Keshavarz

2016-04-01

Full Text Available In this paper, the bearing capacity of strip footings on rock masses has been studied in the seismic case. The stress characteristics or slip line method was used for analysis. The problem was analyzed in the plane strain condition using the Hoek–Brown failure criterion. First, the equilibrium equations along the stress characteristics were obtained and the rock failure criterion was applied. Then, the equations were solved using the finite difference method. A computer code has been provided for analysis. Given the footing and rock parameters, the code can calculate the stress characteristics network and obtain the stress distribution under the footing. The seismic effects have been applied as the horizontal and vertical pseudo-static coefficients. The results of this paper are very close to those of the other studies. The seismic bearing capacity of weightless rock masses can be obtained using the proposed equations and graphs without calculating the whole stress characteristics network.

Summary report of seismic PSA of BWR model plant

International Nuclear Information System (INIS)

1999-05-01

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

Simplified inelastic seismic response analysis of piping system using improved capacity spectrum method

International Nuclear Information System (INIS)

Iijima, Tadashi

2005-01-01

We applied improved capacity spectrum method (ICSM) to a piping system with an asymmetric load-deformation relationship in a piping elbow. The capacity spectrum method can predict an inelastic response by balancing the structural capacity obtained from the load-deformation relationship with the seismic demand defined by an acceleration-displacement response spectrum. The ICSM employs (1) effective damping ratio and period that are based on a statistical methodology, (2) practical procedures necessary to obtain a balance between the structural capacity and the seismic demand. The effective damping ratio and period are defined so as to maximize the probability that predicted response errors lie inside the -10 to 20% range. However, without taking asymmetry into consideration the displacement calculated by using the load-deformation relationship on the stiffer side was 39% larger than that of a time history analysis by a direct integral method. On the other hand, when asymmetry was taken into account, the calculated displacement was only 14% larger than that of a time history analysis. Thus, we verified that the ICSM could predict the inelastic response with errors lying within the -10 to 20% range, by taking into account the asymmetric load-deformation relationship of the piping system. (author)

Preliminary Seismic Response and Fragility Analysis for DACS Cabinet

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

Preliminary Seismic Response and Fragility Analysis for DACS Cabinet

International Nuclear Information System (INIS)

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

2013-01-01

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

Fragility analysis of a seismically-isolated emergency diesel generator

International Nuclear Information System (INIS)

Choun, Young Sun; Choi, In Kil; Ohtori, Yasuki

2005-01-01

The seismic capacity of an Emergency Diesel Generator (EDG) in nuclear power plants influences the seismic safety of the plants significantly. A recent study showed that the increase of the seismic capacity of the EDG could reduce the core damage frequency (CDF) remarkably. It is known that the major failure mode of the EDG is a concrete coning failure due to the pulling out of the anchor bolts. The use of base isolators instead of anchor bolts can increase the seismic capacity of the EDG without any major problems. The fragility curves for a base-isolated EDG should be different from those for a conventional type because the major failure mode of the base-isolated EDG will not be a concrete coning one any more. The governing failure mode of the base-isolated EDG must be the damage of the isolators. This study introduces a fragility evaluation method for an isolated EDG, and evaluates the fragilities for the isolated EDG and compares them with those for the conventional one. Evaluation of the ground motion index is also carried out to determine the governing parameter suitable for representing the seismic responses of the base isolator

Investigation of optimal seismic design methodology for piping systems supported by elasto-plastic dampers. Part 1. Evaluation functions

International Nuclear Information System (INIS)

Ito, Tomohiro; Michiue, Masashi; Fujita, Katsuhisa

2009-01-01

In this study, the optimal seismic design methodology that can consider the structural integrity of not only the piping systems but also elasto-plastic supporting devices is developed. This methodology employs a genetic algorithm and can search the optimal conditions such as the supporting location, capacity and stiffness of the supporting devices. Here, a lead extrusion damper is treated as a typical elasto-plastic damper. Four types of evaluation functions are considered. It is found that the proposed optimal seismic design methodology is very effective and can be applied to the actual seismic design for piping systems supported by elasto-plastic dampers. The effectiveness of the evaluation functions is also clarified. (author)

Use of experience data for DOE seismic evaluations

International Nuclear Information System (INIS)

Barlow, M.W.; Budnitz, R.; Eder, S.J.; Eli, M.W.

1993-01-01

As dictated by DOE Order 5480.28, seismic evaluations of essential systems and components at DOE facilities will be conducted over the next several years. For many of these systems and components, few, if any, seismic requirements applied to the original design, procurement, installation, and maintenance process. Thus the verification of the seismic adequacy of existing systems and components presents a difficult challenge. DOE has undertaken development of the criteria and procedures for these seismic evaluations that will maximize safety benefits in a timely and cost effective manner. As demonstrated in previous applications at DOE facilities and by the experience from the commercial nuclear power industry, use of experience data for these evaluations is the only viable option for most existing systems and components. This paper describes seismic experience data, the needs at DOE facilities, the precedent of application at nuclear power plants and DOE facilities, and the program being put in place for the seismic verification task ahead for DOE

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

International Nuclear Information System (INIS)

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

1981-10-01

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

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.

The Experimental Research on Seismic Capacity of the Envelope Systems with Steel Frame

Science.gov (United States)

Li, Jiuyang; Wang, Bingbing; Li, Hengxu

2017-09-01

In this paper, according to the present application situation of the external envelope systems steel frame in the severe cold region, the stuffed composite wall panels are improved, the flexible connection with the steel frame is designed, the reduced scale specimens are made, the seismic capacity test is made and some indexes of the envelope systems such as bearing capacity, energy consumption and ductility, etc. are compared, which provide reference for the development and application of the steel frame envelope systems.

LANL seismic screening method for existing buildings

International Nuclear Information System (INIS)

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

1997-01-01

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

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.

Seismic risk evaluation for high voltage air insulated substations

International Nuclear Information System (INIS)

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

1997-01-01

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

BUILDING 341 Seismic Evaluation

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-15

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

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)

Lower bound earthquake magnitude for probabilistic seismic hazard evaluation

International Nuclear Information System (INIS)

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

1990-01-01

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

Aspects of the Iea-R1 research reactor seismic evaluation

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1996-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Report of the task group on the seismic behaviour of structures: status report

International Nuclear Information System (INIS)

1997-04-01

In 1995, the CSNI Committee on the Safety of Nuclear Installations) approved a new mandate for PWG-3 and the new title 'Integrity of Components and Structures'. The PWG-3 is assisted by three task groups, one of which is addressing the problem of seismic behavior of structures. Ten topics were identified: engineering characterization of seismic input, site response, soil structure interaction, identification of functions and classification of systems, structures and components, structural response and capacity evaluation (including effects of aging and degradation), component and equipment response and capacity evaluation (including effects of aging and degradation), response and capacity evaluation of distribution systems (piping, cable trays, conduit, HVAC), load combination and acceptance criteria, uncertainties (PSA and margins), plant seismic instrumentation and trip. This report summarizes the seismic issues and activities in various member countries (Canada, Czech Republic, France, Hungary, Italy, Japan, Spain, Switzerland, United Kingdom, United States) and international organizations (IAEA), provides a summary of the important issues that are of collective interest to the group members, and recommends a future programme of work to address these issues

Seismic fragility evaluation of unreinforced masonry walls

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Seismic fragility evaluation of unreinforced masonry walls

International Nuclear Information System (INIS)

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

1991-01-01

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

Optimal Retrofit Scheme for Highway Network under Seismic Hazards

Directory of Open Access Journals (Sweden)

Yongxi Huang

2014-06-01

Full Text Available Many older highway bridges in the United States (US are inadequate for seismic loads and could be severely damaged or collapsed in a relatively small earthquake. According to the most recent American Society of Civil Engineers’ infrastructure report card, one-third of the bridges in the US are rated as structurally deficient and many of these structurally deficient bridges are located in seismic zones. To improve this situation, at-risk bridges must be identified and evaluated and effective retrofitting programs should be in place to reduce their seismic vulnerabilities. In this study, a new retrofit strategy decision scheme for highway bridges under seismic hazards is developed and seamlessly integrate the scenario-based seismic analysis of bridges and the traffic network into the proposed optimization modeling framework. A full spectrum of bridge retrofit strategies is considered based on explicit structural assessment for each seismic damage state. As an empirical case study, the proposed retrofit strategy decision scheme is utilized to evaluate the bridge network in one of the active seismic zones in the US, Charleston, South Carolina. The developed modeling framework, on average, will help increase network throughput traffic capacity by 45% with a cost increase of only $15million for the Mw 5.5 event and increase the capacity fourfold with a cost of only $32m for the Mw 7.0 event.

Seismic re-evaluation of Heavy Water Plant, Kota

International Nuclear Information System (INIS)

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

2003-10-01

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

A quantitative evaluation of seismic margin of typical sodium piping

International Nuclear Information System (INIS)

Morishita, Masaki

1999-05-01

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

Overview on seismic evaluation and retrofitting within JICA Technical Cooperation Project on reduction of seismic risk in Romania

International Nuclear Information System (INIS)

Seki, M.; Vacareanu, R.; Pavel, M.; Lozinca, E.; Cotofana, D.; Chesca, B.; Georgescu, B.; Kaminosono, T.

2007-01-01

The objective of this paper is to give an overview on the seismic evaluation and retrofitting procedures of reinforced concrete buildings within JICA technical cooperation project in Romania. The content of the paper covers a) an outline of the seismic evaluation; history and comparison of Romanian seismic design codes with the Japanese seismic evaluation guidelines, b) an outline of the retrofitting techniques which were transferred from Japan to Romania and structural tests for retrofitting techniques employed in Romania and c) retrofitting details that were used by JICA/NCSRR in the retrofitting design of two vulnerable buildings in Bucharest. The above-mentioned retrofitting projects are now under development of detailed design and therefore, in the near future, refining and improvement of solutions will be performed. (authors)

Seismic re-evaluation of French nuclear power plants

International Nuclear Information System (INIS)

Andrieu, R.

1995-01-01

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

SEISMIC EVALUATION OF EXISTING MID-RISE REINFORCED CONCRETE BUILDINGS ACCORDING TO SPECIFICATION FOR BUILDING STRUCTURES TO BE BUILT IN DISASTER AREAS

Directory of Open Access Journals (Sweden)

Mehmet İNEL

2007-01-01

Full Text Available Over the past several decades, Turkey has been hit by devastating earthquakes and remarkable number of reinforced concrete buildings has been damaged in the high seismicity regions of our country. The aim of this study is to evaluate the seismic performance of the mid-rise reinforced concrete buildings that are major part of building stock of our earthquake-prone country, according to recent Turkish Earthquake Code. 4- and 7-story buildings were selected to represent mid-rise building stock. After determining the structural parameters, each building was subjected to pushover analysis and the capacity curves were obtained. Earthquake performance of each building was determined in the light of their capacity curves according to the recent Turkish Earthquake Code.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

International Nuclear Information System (INIS)

Hu Yuxian

1997-01-01

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

Burar seismic station: evaluation of seismic performance

International Nuclear Information System (INIS)

Ghica, Daniela; Popa, Mihaela

2005-01-01

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

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

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

Pickering seismic safety margin

International Nuclear Information System (INIS)

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

1992-06-01

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

Test on large-scale seismic isolation elements

International Nuclear Information System (INIS)

Mazda, T.; Shiojiri, H.; Oka, Y.; Fujita, T.; Seki, M.

1989-01-01

Demonstration test of seismic isolation elements is considered as one of the most important items in the application of seismic isolation system to fast breeder reactor (FBR) plant. Facilities for testing seismic isolation elements have been built. This paper reports on tests for fullscale laminated rubber bearing and reduced scale models are conducted. From the result of the tests, the laminated rubber bearings turn out to satisfy the specification. Their basic characteristics are confirmed from the tests with fullscale and reduced scale models. The ultimate capacity of the bearings under the condition of ordinary temperature are evaluated

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

Evaluation of Seismic Risk of Siberia Territory

Science.gov (United States)

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

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

Armenian nuclear power plant: US NRC assistance programme for seismic upgrade and safety analysis

International Nuclear Information System (INIS)

Simos, N.; Perkins, K.; Jo, J.; Carew, J.; Ramsey, J.

2003-01-01

This paper summarizes the U.S. Nuclear Regulatory Commission's (US NRC) technical support program activities associated with the Armenian Nuclear Power Plant (ANPP) safety upgrade. The US NRC program, integrated within the overall IAEA-led initiative for safety re-evaluation of the WWER plants, has as its main thrust the technical support to the Armenian Nuclear Regulatory Authority (ANRA) through close collaboration with the scientific staff at Brookhaven National Laboratory (BNL). Several major technical areas of support to ANRA form the basis of the NRC program. These include the seismic re-evaluation and upgrade of the ANPP, safety evaluation of critical systems, and the generation of the Safety Analysis Report (SAR). Specifically, the seismic re-evaluation of the ANPP is part of a broader activity that involves the re-assessment of the seismic hazard at the site, the identification of the Safe Shutdown Equipment at the plant and the evaluation of their seismic capacity, the detailed modeling and analysis of the critical facilities at ANPP, and the generation of the Floor Response Spectra (FRS). Based on the new spectra that incorporate all new findings (hazard, site soil, structure, etc.), the overall capacity of the main structures and the seismic capacity of the critical systems are being re-evaluated. In addition, analyses of critical safe shutdown systems and safe shutdown processes are being performed to ensure both the capabilities of the operating systems and the enhancement of safety due to system upgrades. At present, one of the principal goals of the US NRC's regulatory assistance activities with ANRA is enhancing ANRA's regulatory oversight of high-priority safety issues (both generic and plant-specific) associated with operation of the ANPP. As such, assisting ANRA in understanding and assessing plant-specific seismic and other safety issues associated with the ANPP is a high priority given the ANPP's being located in a seismically active area

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-01

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

An academic program for experience-based seismic evaluation

International Nuclear Information System (INIS)

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

1990-01-01

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

Seismic response of base isolated auxiliary building with age related degradation

International Nuclear Information System (INIS)

Park, Jun Hee; Choun, Young Sun; Choi, In Kil

2012-01-01

The aging of an isolator affects not only the mechanical properties of the isolator but also the dynamic properties of the upper structure, such as the change in stiffness, deformation capacity, load bearing capacity, creep, and damping. Therefore, the seismic response of base isolated structures will change with time. The floor response in the base isolated nuclear power plants (NPPs) can be particularly changed because of the change in stiffness and damping for the isolator. The increased seismic response due to the aging of isolator can cause mechanical problems for many equipment located in the NPPs. Therefore, it is necessary to evaluate the seismic response of base isolated NPPs with age related degradation. In this study, the seismic responses for a base isolated auxiliary building of SHIN KORI 3 and 4 with age related degradation were investigated using a nonlinear time history analysis. Floor response spectrums (FRS) were presented with time for identifying the change in seismic demand under the aging of isolator

Overview of seismic re-evaluation methodologies

International Nuclear Information System (INIS)

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

1993-01-01

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

Seismic evaluation of piping systems using screening criteria

International Nuclear Information System (INIS)

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

1994-01-01

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

Seismic Performance Evaluation of Reinforced Concrete Frames Subjected to Seismic Loads

Science.gov (United States)

Zameeruddin, Mohd.; Sangle, Keshav K.

2017-06-01

Ten storied-3 bays reinforced concrete bare frame designed for gravity loads following the guidelines of IS 456 and IS 13920 for ductility is subjected to seismic loads. The seismic demands on this building were calculated by following IS 1893 for response spectra of 5% damping (for hard soil type). Plastic hinges were assigned to the beam and column at both ends to represent the failure mode, when member yields. Non-linear static (pushover) analysis was performed to evaluate the performance of the building in reference to first (ATC 40), second (FEMA 356) and next-generation (FEMA 440) performance based seismic design procedures. Base shear against top displacement curve of structure, known as pushover curve was obtained for two actions of plastic hinge behavior, force-controlled (brittle) and deformation-controlled (ductile) actions. Lateral deformation corresponding to performance point proves the building capability to sustain a certain level of seismic loads. The failure is represented by a sequence of formation of plastic hinges. Deformation-controlled action of hinges showed that building behaves like strong-column-weak-beam mechanism, whereas force-controlled action showed formation of hinges in the column. The study aims to understand the first, second and next generation performance based design procedure in prediction of actual building responses and their conservatism into the acceptance criteria.

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

Science.gov (United States)

2013-02-26

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

Potential seismic structural failure modes associated with the Zion Nuclear Plant. Seismic safety margins research program (Phase I). Project VI. Fragilities

International Nuclear Information System (INIS)

1979-10-01

The Zion 1 and 2 Nuclear Power Plant consists of a number of structures. The most important of these from the viewpoint of safety are the containment buildings, the auxiliary building, the turbine building, and the crib house (or intake structure). The evaluation of the potential seismic failure modes and determination of the ultimate seismic capacity of the structures is a complex undertaking which will require a large number of detailed calculations. As the first step in this evaluation, a number of potential modes of structural failure have been determined and are discussed. The report is principally directed towards seismically induced failure of structures. To some extent, modes involving soil foundation failures are discussed in so far as they affect the buildings. However, failure modes involving soil liquefaction, surface faulting, tsunamis, etc., are considered outside the scope of this evaluation

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

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Yoon, Doo Byung

2003-06-01

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

On development and improvement of evaluation techniques for seismic ground motion

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

Preliminary evaluation of the seismic hazard at Cernavoda NPP site

International Nuclear Information System (INIS)

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

2001-01-01

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

Re-evaluation and updating of the seismic hazard of Lebanon

Science.gov (United States)

Huijer, Carla; Harajli, Mohamed; Sadek, Salah

2016-01-01

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

Seismic structural fragility investigation for the San Onofre Nuclear Generating Station, Unit 1 (Project I); SONGS-1 AFWS Project

International Nuclear Information System (INIS)

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

1982-04-01

An evaluation of the seismic capacities of several of the San Onofre Nuclear Generating Station, Unit 1 (SONGS-1) structures was conducted to determine input to the overall probabilistic methodology developed by Lawrence Livermore National Laboratory. Seismic structural fragilities to be used as input consist of median seismic capacities and their variabilities due to randomness and uncertainty. Potential failure modes were identified for each of the SONGS-1 structures included in this study by establishing the seismic load-paths and comparing expected load distributions to available capacities for the elements of each load-path. Particular attention was given to possible weak links and details. The more likely failure modes were screened for more detailed investigation

Evaluation of seismic hazards for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2002-01-01

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

Evaluation and assessment of nuclear power plant seismic methodology

International Nuclear Information System (INIS)

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

1977-01-01

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

Evaluation and assessment of nuclear power plant seismic methodology

Energy Technology Data Exchange (ETDEWEB)

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

1977-03-01

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

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

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

1989-01-01

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

Seismic evaluation of safety systems at the Savannah River reactors

International Nuclear Information System (INIS)

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

1989-01-01

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

Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

Energy Technology Data Exchange (ETDEWEB)

Choun, Young Sun; Park, Jun Hee [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

International Nuclear Information System (INIS)

Choun, Young Sun; Park, Jun Hee

2015-01-01

Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers

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

Science.gov (United States)

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

2018-03-01

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

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)

International symposium on seismic evaluation of existing nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-01

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

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

International Nuclear Information System (INIS)

Soares, W.A.

1983-03-01

An overall view of the subjects involved in the determination of the site standard seismic motion to a nuclear facility is presented. The main topics discussed are: basic priciples of seismic instrumentation; dynamic and spectral concepts; design earthquakes definitions; fundamentals of seismology; empirical curves developed from prior seismic data; avalable methodologies and recommended procedures to evaluate the site standard seismic motion. (E.G.) [pt

Comparison of evaluation guidelines for life-safety seismic hazards

International Nuclear Information System (INIS)

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

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-08-15

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

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

International Nuclear Information System (INIS)

Ohtori, Yasuki; Kanatani, Mamoru

2006-01-01

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

Seismic evaluation of nuclear installations

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1997-01-01

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

Reliability of lifeline networks under seismic hazard

International Nuclear Information System (INIS)

Selcuk, A. Sevtap; Yuecemen, M. Semih

1999-01-01

Lifelines, such as pipelines, transportation, communication and power transmission systems, are networks which extend spatially over large geographical regions. The quantification of the reliability (survival probability) of a lifeline under seismic threat requires attention, as the proper functioning of these systems during or after a destructive earthquake is vital. In this study, a lifeline is idealized as an equivalent network with the capacity of its elements being random and spatially correlated and a comprehensive probabilistic model for the assessment of the reliability of lifelines under earthquake loads is developed. The seismic hazard that the network is exposed to is described by a probability distribution derived by using the past earthquake occurrence data. The seismic hazard analysis is based on the 'classical' seismic hazard analysis model with some modifications. An efficient algorithm developed by Yoo and Deo (Yoo YB, Deo N. A comparison of algorithms for terminal pair reliability. IEEE Transactions on Reliability 1988; 37: 210-215) is utilized for the evaluation of the network reliability. This algorithm eliminates the CPU time and memory capacity problems for large networks. A comprehensive computer program, called LIFEPACK is coded in Fortran language in order to carry out the numerical computations. Two detailed case studies are presented to show the implementation of the proposed model

Technical evaluation of seismic qualification of safety-related equipment

Energy Technology Data Exchange (ETDEWEB)

Cho, Yang Hui; Park, Heong Gee; Park, Yeong Seok [Univ. of Incheon, Incheon (Korea, Republic of)

1994-04-15

This study is purposed to evaluate the technical acceptability of the procedures and techniques of seismic qualifications which were performed for the YGN 3 and 4 safety-related equipment.This study is also targeted to suggest a systematized technical procedure guide for the effective performance and review of the seismic qualification, which reflects the most up-to-date licensing requirements and state-of the-art.

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

Directory of Open Access Journals (Sweden)

C. V. Christova

2000-06-01

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

Seismic fragility analyses of nuclear power plant structures based on the recorded earthquake data in Korea

International Nuclear Information System (INIS)

Joe, Yang Hee; Cho, Sung Gook

2003-01-01

This paper briefly introduces an improved method for evaluating seismic fragilities of components of nuclear power plants in Korea. Engineering characteristics of small magnitude earthquake spectra recorded in the Korean peninsula during the last several years are also discussed in this paper. For the purpose of evaluating the effects of the recorded earthquake on the seismic fragilities of Korean nuclear power plant structures, several cases of comparative studies have been performed. The study results show that seismic fragility analysis based on the Newmark's spectra in Korea might over-estimate the seismic capacities of Korean facilities. (author)

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)

Functional Capacity Evaluation Research : Report from the Third International Functional Capacity Evaluation Research Meeting

NARCIS (Netherlands)

James, C. L.; Reneman, M. F.; Gross, D. P.

Introduction Functional capacity evaluations are an important component of many occupational rehabilitation programs and can play a role in facilitating reintegration to work thus improving health and disability outcomes. The field of functional capacity evaluation (FCE) research has continued to

Seismic performance evaluation of an infilled rocking wall frame structure through quasi-static cyclic testing

Science.gov (United States)

Pan, Peng; Wu, Shoujun; Wang, Haishen; Nie, Xin

2018-04-01

Earthquake investigations have illustrated that even code-compliant reinforced concrete frames may suffer from soft-story mechanism. This damage mode results in poor ductility and limited energy dissipation. Continuous components offer alternatives that may avoid such failures. A novel infilled rocking wall frame system is proposed that takes advantage of continuous component and rocking characteristics. Previous studies have investigated similar systems that combine a reinforced concrete frame and a wall with rocking behavior used. However, a large-scale experimental study of a reinforced concrete frame combined with a rocking wall has not been reported. In this study, a seismic performance evaluation of the newly proposed infilled rocking wall frame structure was conducted through quasi-static cyclic testing. Critical joints were designed and verified. Numerical models were established and calibrated to estimate frame shear forces. The results evaluation demonstrate that an infilled rocking wall frame can effectively avoid soft-story mechanisms. Capacity and initial stiffness are greatly improved and self-centering behavior is achieved with the help of the infilled rocking wall. Drift distribution becomes more uniform with height. Concrete cracks and damage occurs in desired areas. The infilled rocking wall frame offers a promising approach to achieving seismic resilience.

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

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

International Nuclear Information System (INIS)

2003-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-08-01

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

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.)

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

International Nuclear Information System (INIS)

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

2002-05-01

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

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

Directory of Open Access Journals (Sweden)

Chongwen Jiang

2017-01-01

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

Seismic assessment of Kozloduy WWER-440, model 230 nuclear power plant

International Nuclear Information System (INIS)

Monette, P.; Baltus, R.; Yanev, P.; Campbell, R.

1993-01-01

A preliminary report is given of the findings of an IAEA sponsored walkdown of the WWER-440, model 230 at Kozloduy, in May 1991. The scope of the IAEA mission was to determine the lower bound seismic capacity of the plant and to make recommendations for improvements to increase the earthquake resistance. The methodology utilized in the assessment is that used for evaluation of the seismic margin in U.S. nuclear power plants subjected to earthquakes beyond their design basis. Included in the assessment is the establishment of a safe shutdown path which would include the capacity to mitigate a small break in the primary system, performance of a walkdown of the safe shutdown path and calculation of the high-confidence-of-low-probability-of-failure (HCLPF) of the safe shutdown path. Excluding system design deficiency relative to U.S. and Western Europe standards, it was found that the plant has many seismic vulnerabilities similar to those that existed in many of the U.S. plants prior to about 1979 when the Systematic Evaluation Program was initiated. (Z.S.) 1 tab., 1 fig

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

International Nuclear Information System (INIS)

Soares, W.A.

1983-01-01

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

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

Indian Academy of Sciences (India)

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

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

Science.gov (United States)

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

2016-04-01

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

Integrated software system for seismic evaluation of nuclear power plant structures

International Nuclear Information System (INIS)

Xu, J.; Graves, H.L.

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

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

International Nuclear Information System (INIS)

Ile, Nicolas; Frau, Alberto

2017-01-01

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

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

International Nuclear Information System (INIS)

Stevenson, J.D.

1995-01-01

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

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

International Nuclear Information System (INIS)

Podrouzek, P.

1997-01-01

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

Seismic qualification of a commercial grade emergency diesel generator system in high seismic zones

International Nuclear Information System (INIS)

Khan, Mohsin R.; Chen, Wayne W.H.; Chu, Winnie S.

2004-01-01

The paper presents the seismic qualification of a commercially procured emergency diesel generator (EDG) system for use in a nuclear power plant. Response spectrum analyses of finite element models, validated using in situ vibration test data, were performed to qualify the skid and floor mounted mechanical components whose functional capacity and structural integrity can be analyzed. Time history analyses of these models were also performed to obtain the amplified response spectra for seismic testing of small valves, electrical and electro-mechanical components whose functional capacity can not be analyzed to establish the seismic qualification. The operational loads were obtained by in-plant vibration monitoring. Full scale shake table testing was performed for auxiliary electrical cabinets. It is concluded that with some minor structural modifications, a commercial grade EDG system can be qualified for safety-related applications in nuclear power plants located in high seismic zones. (author)

International symposium on seismic evaluation of existing nuclear facilities

International Nuclear Information System (INIS)

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

2003-10-01

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

Development of Technology for Structural Integrity Evaluation

International Nuclear Information System (INIS)

Choun, Young Sun; Choi, I. K.; Kim, M. K. and others

2005-03-01

The purpose of this study is a development of seismic safety and structural integrity evaluation method of the structure in the Nuclear Power plant (NPP). The purpose of 1st sub-Topic is the development and improvement of the seismic safety evaluation methodology for the Nuclear Power Plant structures and safety related equipment. The purpose of 2nd sub-topic is the increasing of structure and equipment seismic capacity through the reducing of seismic force. The purpose of 3rd sub-topic is the development of 3-D nonlinear finite element analysis program for prestressed concrete containment building. The last purpose if the evaluation of the failure mechanism of containment structure and structure capacity and the assessment of integrity of containment through the of leakage test. As a result of this research, there are many research results were produced. The scenario earthquake developing method was developed and the effect of the structures and equipment was analyzed. The effectiveness of isolation system was determined and optimum isolation systems for each equipment were selected. The NUCAS-3D program for the 3 dimensional numerical analysis of containment building using the embedded tendon element and rebar element was developed. The tension behavior of containment building was examined and the leakage rate of the concrete crack was determined. The results of this research can be successfully used for many fields of integrity of NPP site. It can be used for development of design earthquake for the seismic design and safety evaluation and establishment of seismic safety evaluation program and seismic capacity improvement program for existing NPP. In case of seismic isolation part, it can be used for the application to the selection of optimum isolation devices for equipment isolation and to the effective evaluation of each seismic isolation devices. In containment analysis part, it can be used for ultimate pressure capacity evaluation of prestressed concrete

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

Evaluation of seismic criteria used in design of INEL facilities

International Nuclear Information System (INIS)

Young, G.A.

1977-01-01

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

Seismic evaluation of the Mors Dome

International Nuclear Information System (INIS)

Kreitz, E.

1982-01-01

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

Rethinking ASME III seismic analysis for piping operability evaluations

International Nuclear Information System (INIS)

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

1994-01-01

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

Synthesis of analytical and experimental data, capacity evaluation

International Nuclear Information System (INIS)

Lin Chiwenn

2001-01-01

This part of the presentation deals with the synthesis of analytical and experimental data and capacity evaluation. First, a typical test flow diagram will be discussed to identify key aspects of the test program where analysis is to be performed. Next, actual component test and analysis programs will be presented to illustrate some important parameters to be considered in the modelling process. Then, two combined test and analysis projects will be reviewed to demonstrate the potential use of substructuring in the model testing to reduce the size of the model to be tested. This will be followed by an inelastic response spectral reactor coolant loop analysis, which was used to study a high level seismic test conducted for a PWR reactor coolant system. The potential use of an improved impact calculation method will be discussed after that. As a closure to the test and analysis synthesis process, a reactor internal qualification process will be discussed. Finally, capacity evaluation will be discussed, following the requirements of ASME section III code for class 1 pressure vessel, class 1 piping which includes the reactor coolant loop piping, and the reactor internals. The subsections included in this part of presentation which cover the above mentioned subjects: typical component test and analysis results; combined test and analysis process; a simplified inelastic response spectral; analysis of reactor coolant loop; an improved impact analysis methodology; reactor coolant system and core internal qualification process; ASME section III code, design by analysis of class 1 pressure vessel; design by analysis of class 1 piping; SME section III code, design by analysis of reactor core internals

A progressive methodology for seismic safety evaluation of gravity dams

International Nuclear Information System (INIS)

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

1995-01-01

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

Seismic test for safety evaluation of low level radioactive wastes containers

International Nuclear Information System (INIS)

Ohoka, Makoto; Horikiri, Morito

1998-08-01

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

Seismic evaluation of buildings in the Eastern and Central United States

International Nuclear Information System (INIS)

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

1991-01-01

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

Post-Earthquake Traffic Capacity of Modern Bridges in California

Science.gov (United States)

2010-03-01

Evaluation of the capacity of a bridge to carry self-weight and traffic loads after an earthquake is essential for a : safe and timely re-opening of the bridge. In California, modern highway bridges designed using the Caltrans : Seismic Design Criter...

Evaluation of seismic hazard at the northwestern part of Egypt

Science.gov (United States)

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

2016-01-01

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

Wind/seismic comparison for upgrading existing structures

International Nuclear Information System (INIS)

Giller, R.A.

1989-01-01

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

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%

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

International Nuclear Information System (INIS)

2015-06-01

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

Seismic fragility analyses of nuclear power plant structures based on the recorded earthquake data in Korea

International Nuclear Information System (INIS)

Cho, Sung Gook; Joe, Yang Hee

2005-01-01

By nature, the seismic fragility analysis results will be considerably affected by the statistical data of design information and site-dependent ground motions. The engineering characteristics of small magnitude earthquake spectra recorded in the Korean peninsula during the last several years are analyzed in this paper. An improved method of seismic fragility analysis is evaluated by comparative analyses to verify its efficiency for practical application to nuclear power plant structures. The effects of the recorded earthquake on the seismic fragilities of Korean nuclear power plant structures are also evaluated from the comparative studies. Observing the obtained results, the proposed method is more efficient for the multi-modes structures. The case study results show that seismic fragility analysis based on the Newmark's spectra in Korea might over-estimate the seismic capacities of Korean facilities

Seismic fragility analyses of nuclear power plant structures based on the recorded earthquake data in Korea

Energy Technology Data Exchange (ETDEWEB)

Cho, Sung Gook [Department of Civil and Environmental System Engineering, University of Incheon, 177 Dohwa-dong, Nam-gu, Incheon 402-749 (Korea, Republic of)]. E-mail: sgcho@incheon.ac.kr; Joe, Yang Hee [Department of Civil and Environmental System Engineering, University of Incheon, 177 Dohwa-dong, Nam-gu, Incheon 402-749 (Korea, Republic of)

2005-08-01

By nature, the seismic fragility analysis results will be considerably affected by the statistical data of design information and site-dependent ground motions. The engineering characteristics of small magnitude earthquake spectra recorded in the Korean peninsula during the last several years are analyzed in this paper. An improved method of seismic fragility analysis is evaluated by comparative analyses to verify its efficiency for practical application to nuclear power plant structures. The effects of the recorded earthquake on the seismic fragilities of Korean nuclear power plant structures are also evaluated from the comparative studies. Observing the obtained results, the proposed method is more efficient for the multi-modes structures. The case study results show that seismic fragility analysis based on the Newmark's spectra in Korea might over-estimate the seismic capacities of Korean facilities.

Seismic qualification of motor operated valves - alternate approach

International Nuclear Information System (INIS)

Bruck, P.M.; Eissa, M.A.

1998-01-01

This paper presents a potential alternate method for determining operating capacity of motor-operated valves subjected to seismic and other applicable loadings. As a result of programs at nuclear facilities to ensure the operational capability of MOVs (under NRC GL89-10), extensive analytical focus to develop the structural capability of valves has ensued. In the past, seismic qualification of valves typically addressed the strength of the topwork structure to resist inertial loading from excitation of the large valve actuator mass. These evaluations paid little or no consideration to the loading resulting from valve closing forces. The focus of the recent efforts is to develop the maximum operational capability of the valve, in terms of thrust, with consideration of seismic and other services loading as applicable. The alternate method outlined in this paper presents a series of thrust capacity curves, with reduction factors for seismic loading which can be applied and developed to determine safe thrust loadings without performing extensive analytical effort. A similar approach was put forward by the SQUG GIP approach to MOVs to ensure the safe operation of valves based on past earthquake experience. However, the GIP approach cannot be used to determine safe operational loads and thus has limited use in the necessary analysis required for GL89-10 programs at nuclear facilities. (orig.)

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

International Nuclear Information System (INIS)

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

1993-02-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Probabilistic seismic hazards: Guidelines and constraints in evaluating results

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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.)

Overview of seismic margin insights gained from seismic PRA results

International Nuclear Information System (INIS)

Kennedy, R.P.; Sues, R.H.; Campbell, R.D.

1986-01-01

This paper presents the findings of a study conducted under NRC and EPRI sponsorship in which published seismic PRAs were reviewed in order to gain insight to the seismic margins inherent in existing nuclear plants. The approach taken was to examine the fragilities of those components which have been found to be dominant contributors to seismic risk at plants in low-to-moderate seismic regions (SSE levels between 0.12g and 0.25g). It is concluded that there is significant margin inherent in the capacity of most critical components above the plant design basis. For ground motions less than about 0.3g, the predominant sources of seismic risk are loss of offsite power coupled with random failure of the emergency diesels, non-recoverable circuit breaker trip due to relay chatter, unanchored equipment, unreinforced non-load bearing block walls, vertical water storage tanks, systems interactions and possibly soil liquefaction. Recommendations as to which components should be reviewed in seismic margin studies for margin earthquakes less than 0.3g, between 0.3g and 0.5g, and greater than 0.5g, developed by the NRC expert panel on the quantification of seismic margins (based on the review of past PRA data, earthquake experience data, and their own personal experience) are presented

Risk management considerations for seismic upgrading of an older facility for short-term residue stabilization

International Nuclear Information System (INIS)

Additon, S.L.; Peregoy, W.L.; Foppe, T.L.

1999-01-01

Building 707 and its addition, Building 707A, were selected, after the production mission of Rocky Flats was terminated a few years ago, to stabilize many of the plutonium residues remaining at the site by 2002. The facility had undergone substantial safety improvements to its safety systems and conduct of operations for resumption of plutonium operations in the early 1990s and appeared ideally suited for this new mission to support accelerated Site closure. During development of a new authorization basis, a seismic evaluation was performed. This evaluation addressed an unanalyzed expansion joint and suspect connection details for the precast concrete tilt-up construction and concluded that the seismic capacity of the facility is less than half of that determined by previous analysis. Further, potential seismic interaction was identified between a collapsing Building 707 and the seismically upgraded Building 707A, possibly causing the partial collapse of the latter. Both the operating contractor and the Department of Energy sought a sound technical basis for deciding how to proceed. This paper addresses the risks of the as-is facility and possible benefits of upgrades to support a decision on whether to upgrade the seismic capacity of Building 707, accept the risk of the as-is facility for its short remaining mission, or relocate critical stabilization missions. The paper also addresses the Department of Energy's policy on natural phenomena

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.

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

Indian Academy of Sciences (India)

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

Seismic risk assessment of Trani's Cathedral bell tower in Apulia, Italy

Science.gov (United States)

Diaferio, Mariella; Foti, Dora

2017-09-01

The present paper deals with the evaluation of the seismic vulnerability of slender historical buildings; these structures, in fact, may manifest a high risk with respect to seismic actions as usually they have been designed to resist to gravitational loads only, and are characterized by a high flexibility. To evaluate this behavior, the bell tower of the Trani's Cathedral is investigated. The tower is 57 m tall and is characterized by an unusual building typology, i.e., the walls are composed of a concrete core coupled with external masonry stones. The dynamic parameters and the mechanical properties of the tower have been evaluated on the basis of an extensive experimental campaign that made use of ambient vibration tests and ground penetrating radar tests. Such data have been utilized to calibrate a numerical model of the examined tower. A linear static analysis, a dynamic analysis and a nonlinear static analysis have been carried out on such model to evaluate the displacement capacity of the tower and the seismic risk assessment in accordance with the Italian guidelines.

Seismic qualification of equipment

International Nuclear Information System (INIS)

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

1983-03-01

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

Seismic evaluation of a diesel generator system at the Savannah River Site using earthquake experience data

International Nuclear Information System (INIS)

Griffin, M.J.; Tong, Wen H.; Rawls, G.B.

1990-01-01

New equipment and systems have been seismically qualified traditionally by either two methods, testing or analysis. Testing programs are generally expensive and their input loadings are conservative. It is generally recognized that standard seismic analysis techniques produce conservative results. Seismic loads and response levels for equipment that are typically calculated exceed the values actually experienced in earthquakes. An alternate method for demonstrating the seismic adequacy of equipment has been developed which is based on conclusions derived from studying the performance of equipment that has been subjected to actual earthquake excitations. The conclusion reached from earthquake experience data is that damage or malfunction to most types of equipment subjected to earthquakes is less than that predicted by traditional testing and analysis techniques. The use of conclusions derived from experience data provides a realistic approach in assessing the seismic ruggedness of equipment. By recognizing the inherently higher capacity that exists in specific classes of equipment, commercial ''off-the-shelf'' equipment can be procured and qualified without the need to perform expensive modifications to meet requirements imposed by traditional conservative qualification analyses. This paper will present the seismic experience data methodology applied to demonstrate the seismic adequacy of several commercially supplied 800KW diesel powered engine driven generator sets with peripheral support components installed at the Savannah River Site (SRS)

Seismic evaluation of Tank 241C106 in support of retrieval activities

International Nuclear Information System (INIS)

Wallace, D.A.

1994-01-01

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

Seismic re-evaluation of the Tarapur atomic power plants 1 and 2

International Nuclear Information System (INIS)

Ingole, S.M.; Kumar, B.S.; Gupta, S.; Singh, U.P.; Giridhar, K.; Bhawsar, S.D.; Samota, A.; Chhatre, A.G.; Dixit, K.B.; Bhardwaj, S.A.

2004-01-01

Two Boiling Water Reactors (BWR) of 210 MWe each at Tarapur Atomic Power Station, Units-1 and 2 (TAPS-1-2) were commissioned in the year 1969. The safety related civil structures at TAPS had been designed for a seismic coefficient of 0.2 g and other structures for 0.1 g. The work of seismic re-evaluation of the TAPS-1-2 has been taken up in the year 2002. As two new Pressurized Heavy Water Reactor (PHWR) plants of 540 MWe each, Tarapur Atomic Power Project Units-3 and 4 (TAPP-3-4), are coming up in the vicinity of TAPS-1-2, detailed geological and seismological studies of the area around TAPS-1-2 are available. The same free-field ground motion as generated for TAPP-3-4 has been used for TAPS-1-2. The seismic re-evaluation of the plant has been performed as per the procedure given in IAEA, Safety Reports Series entitled 'Seismic Evaluation of Existing Nuclear Power Plants', and meeting the various codes and standards, viz., ASME, ASCE, IEEE standards etc. The Safety Systems (SS) and Safety Support Systems (SSS) are qualified by adopting detailed analysis and testing methods. The equipment in the SS and SSS have been qualified by conducting a walk-down as per the procedure given in Generic Implementation Procedure, Dept. of Energy (GIP--DOE), USA. The safety systems include the systems required for safe shutdown of the plant, one chain of decay heat removal and containment of activity. The safety support systems viz., Electrical, Instrumentation and Control and systems other than SS and SSS have been qualified by limited analysis, testing and mostly by following the procedure of walk-down. The paper brings out the details of the work accomplished during seismic re-evaluation of the two units of BWR at Tarapur. (authors)

Reducing Seismic Hazard and Building Capacity Through International Cooperation

Science.gov (United States)

Vergino, E. S.; Arakelyan, A.; Babayan, H.; Durgaryan, R.; Elashvili, M.; Godoladze, T.; Javakhishvili, Z.; Kalogeras, I.; Karakhanyan, A.; Martin, R. J.; Yetirmishli, G.

2012-12-01

During the last 50 years, the Caucasus, Central Asia and the Caspian Sea regions have experienced several devastating earthquakes. While each country in the region has worked with its neighbors on small, ad-hoc projects to improve preparedness, deeply ingrained political and ethnic rivalries, and severely stressed economies have severely hindered sustained regional cooperation. Future damaging earthquakes are inevitable and without proper planning the negative impact on public safety, security, economics and stability in these regions will be devastating. We have, through twelve years of international scientific cooperation, focused on the development of an expanded skill base and infrastructure, through the installation of new, modern, digital seismic monitoring networks, building of historic databases, sharing seismic, geologic and geophysical data, conducting joint scientific investigations utilizing the new digital data and applying modern techniques, as well as the development of regional hazard models that the scientists of the region share with their governments and use to advise them on the best ways to mitigate the impact of a damaging earthquake. We have established specialized regional scientific task-force teams who can carry out seismological, geological and engineering studies in the epicentral zone, including the collection of new scientific data, for better understanding of seismic and geodynamic processes as well to provide emergency support in crisis and post-crisis situations in the Southern Caucasus countries. "Secrecy" in crisis and post-crisis situations in the former Soviet Union countries, as well as political instabilities, led to an absence of seismic risk reduction and prevention measures as well as little to no training of scientific-technical personnel who could take action in emergency situations. There were few opportunities for the development of a next generation of scientific experts, thus we have placed emphasis on the inclusion

Ground motion input in seismic evaluation studies

International Nuclear Information System (INIS)

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

1996-07-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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)

Structural failure modes in vertical tanks: reinforcement evaluation and solutions

International Nuclear Information System (INIS)

Alcantud Abellan, M.; Orden Martinez, A.

1995-01-01

Vertical storage tanks are essential components in the safety of nuclear plant systems. It has been shown that the traditional method of analysing seismic loads is not conservative, as it does not take account of the interaction between fluid and tank structure. This paper identifies different possible structural failure modes in tanks due to seismic load, and methods devised by various authors to evaluate tank structure capacity under different failure modes. These methods are based on experimental data relating to the structural behaviour of tanks during actual seismic events, tests, and theoretical analyses. The paper describes the problems of these structures under seismic loads in nuclear plants. It proposes solutions to the main structural problem, tank anchorage, for which the re-evaluation of the anchorage capacity is required, using methods (finite element) less conservative than those proposed by other authors. Also proposed is the local reinforcement of anchorages to increase their capacity. (Author) 4 refs

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)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-01-01

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

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

International Nuclear Information System (INIS)

2002-01-01

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

Seismic fatigue life evaluation of mechanical structures using energy balance equation

International Nuclear Information System (INIS)

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

2009-01-01

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

Seismic fragility analysis of structural components for HFBR facilities

International Nuclear Information System (INIS)

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

1992-01-01

The paper presents a summary of recently completed seismic fragility analyses of the HFBR facilities. Based on a detailed review of past PRA studies, various refinements were made regarding the strength and ductility evaluation of structural components. Available laboratory test data were analysed to evaluate the formulations used to predict the ultimate strength and deformation capacities of steel, reinforced concrete and masonry structures. The biasness and uncertainties were evaluated within the framework of the fragility evaluation methods widely accepted in the nuclear industry. A few examples of fragility calculations are also included to illustrate the use of the presented formulations

Seismic Evaluation of Hydrocarbon Saturation in Deep-Water Reservoirs

Energy Technology Data Exchange (ETDEWEB)

Michael Batzle

2006-04-30

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

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

Energy Technology Data Exchange (ETDEWEB)

Ha, Jeong Gon, E-mail: jgha87@kaist.ac.kr; Kim, Dong-Soo, E-mail: dskim@kaist.ac.kr

2014-10-01

Highlights: • A series of dynamic centrifuge tests were performed for NPP structure to investigate the soil–foundation-structure interaction with various soil conditions from loose sand to weathered rock. • SFSI phenomena for NPP structure were observed directly using experimental method. • Effect of the soil stiffness and nonlinear characteristics on SFSI was estimated. • There are comparisons of the control motions for seismic design of a NPP structure. • Subsoil condition, earthquake intensity and control motion affected to seismic load. - Abstract: To evaluate the earthquake loads for the seismic design of a nuclear containment structure, it is necessary to consider the soil–foundation-structure interaction (SFSI) due to their interdependent behavior. Especially, understanding the effects of soil stiffness under the structure and the location of control motion to SFSI are very important. Motivated by these requirements, a series of dynamic centrifuge tests were performed with various soil conditions from loose sand to weathered rock (WR), as well as different seismic intensities for the bedrock motion. The different amplification characteristics in peak-accelerations profile and effects of soil-nonlinearity in response spectrum were observed. The dynamic behaviors were compared between surface of free-field and foundation of the structure for the evaluation of the control motion for seismic design. It was found that dynamic centrifuge test has potentials to estimate the seismic load considering SFSI.

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

International Nuclear Information System (INIS)

Ha, Jeong Gon; Kim, Dong-Soo

2014-01-01

Highlights: • A series of dynamic centrifuge tests were performed for NPP structure to investigate the soil–foundation-structure interaction with various soil conditions from loose sand to weathered rock. • SFSI phenomena for NPP structure were observed directly using experimental method. • Effect of the soil stiffness and nonlinear characteristics on SFSI was estimated. • There are comparisons of the control motions for seismic design of a NPP structure. • Subsoil condition, earthquake intensity and control motion affected to seismic load. - Abstract: To evaluate the earthquake loads for the seismic design of a nuclear containment structure, it is necessary to consider the soil–foundation-structure interaction (SFSI) due to their interdependent behavior. Especially, understanding the effects of soil stiffness under the structure and the location of control motion to SFSI are very important. Motivated by these requirements, a series of dynamic centrifuge tests were performed with various soil conditions from loose sand to weathered rock (WR), as well as different seismic intensities for the bedrock motion. The different amplification characteristics in peak-accelerations profile and effects of soil-nonlinearity in response spectrum were observed. The dynamic behaviors were compared between surface of free-field and foundation of the structure for the evaluation of the control motion for seismic design. It was found that dynamic centrifuge test has potentials to estimate the seismic load considering SFSI

Structural evaluation of the 2736Z Building for seismic loads

International Nuclear Information System (INIS)

Giller, R.A.

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Factors controlling strength of structures, and anticipated overstrength for seismic load conditions

International Nuclear Information System (INIS)

Singh, A.K.

1985-01-01

This paper discusses how the safe shutdown earthquake level, the ratio of operating basis earthquake to safe shutdown earthquake level, the shape of the earthquake spectra and the modeling of the structure affect the seismic overstrength of structures. The relationship between actual mean strength and the minimum specified strength of concrete and structural steel is also presented. The paper identifies which concrete and steel structures are generally sized for earthquake loads and which are generally sized for other factors, e.g., tornado missiles, loss of coolant accident pressure loads, equipment laydown loads and radiation shielding. The results of a study evaluating the mean ultimate capacity of a pressurized water reactor containment are presented to show that in terms of a ground motion, the seismic capacity may be four to five times the design safe shutdown earthquake level

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

OpenAIRE

Kamagata, S.; Takewaki, Izuru

2015-01-01

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

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

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

Seismic Response and Performance Evaluation of Self-Centering LRB Isolators Installed on the CBF Building under NF Ground Motions

Directory of Open Access Journals (Sweden)

Junwon Seo

2016-01-01

Full Text Available This paper mainly treats the seismic behavior of lead-rubber bearing (LRB isolation systems with superealstic shape memory alloy (SMA bending bars functioning as damper and self-centering devices. The conventional LRB isolators that are usually installed at the column bases supply extra flexibility to the centrically braced frame (CBF building with a view to elongate its vibration period, and thus make a contribution to mitigating seismic acceleration transferred from ground to structure. However, these base isolation systems are somehow susceptible to shear failure due to the lack of lateral resistance. In the construction site, they have been used to be integrated with displacement control dampers additionally withstanding lateral seismic forces. For this motivation, LRB isolation systems equipped with superelastic SMA bending bars, which possess not only excellent energy dissipation but also outstanding recentering capability, are proposed in this study. These reinforced and recentering LRB base isolators are modeled as nonlinear component springs, and then assigned into the bases of 2D frame models used for numerical simulation. Their seismic performance and capacity in the base-isolated frame building can be evaluated through nonlinear dynamic analyses conducted with historic ground motion data. After comparative study with analyses results, it is clearly shown that 2D frame models with proposed LRB isolators generally have smaller maximum displacements than those with conventional LRB isolators. Furthermore, the LRB isolation systems with superelastic SMA bending bars effectively reduce residual displacement as compared to those with steel bending bars because they provide more flexibility and recentering force to the entire building structure.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Science.gov (United States)

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

2017-12-01

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

Seismic fragility analysis of a nuclear building based on probabilistic seismic hazard assessment and soil-structure interaction analysis

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, R.; Ni, S.; Chen, R.; Han, X.M. [CANDU Energy Inc, Mississauga, Ontario (Canada); Mullin, D. [New Brunswick Power, Point Lepreau, New Brunswick (Canada)

2016-09-15

Seismic fragility analyses are conducted as part of seismic probabilistic safety assessment (SPSA) for nuclear facilities. Probabilistic seismic hazard assessment (PSHA) has been undertaken for a nuclear power plant in eastern Canada. Uniform Hazard Spectra (UHS), obtained from the PSHA, is characterized by high frequency content which differs from the original plant design basis earthquake spectral shape. Seismic fragility calculations for the service building of a CANDU 6 nuclear power plant suggests that the high frequency effects of the UHS can be mitigated through site response analysis with site specific geological conditions and state-of-the-art soil-structure interaction analysis. In this paper, it is shown that by performing a detailed seismic analysis using the latest technology, the conservatism embedded in the original seismic design can be quantified and the seismic capacity of the building in terms of High Confidence of Low Probability of Failure (HCLPF) can be improved. (author)

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

International Nuclear Information System (INIS)

2016-07-01

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

Seismic Applications of Energy Dampers

OpenAIRE

Shambhu Sinha

2004-01-01

Damping devices based on the operating principle of high velocity fluid flow through orifices have found numerous applications in the shock and vibration isolation of aerospace and defence systems. The study aims to investigate the feasibility of using energy dissipating fluid viscous dampers in structures to protect against seismic loads and to prove analytically and  experimentally that fluid viscous dampers can improve the seismic capacity of a structure by reducing damage and displacement...

Evaluation of seismic damage to bridges and highway systems in Shelby County, Tennessee

Science.gov (United States)

Jernigan, John Bailey

Past earthquakes have demonstrated that bridges are one of the most vulnerable components of highway transportation systems. In addition to bridges, roadways may also be subject to damage, particularly in an area prone to earthquake-induced liquefaction. As a consequence, the highway transportation systems after an earthquake might be impaired and the post-earthquake emergency response might be compromised. Furthermore, the impact on the regional economy might be very significant from the damage to highway systems. Since highway transportation systems are critical lifelines for people living in an urban area, it is important to evaluate the vulnerability of bridges and highway systems in earthquake-prone regions. Memphis and Shelby County, Tennessee are located close to the southwestern segment of the New Madrid seismic zone (NMSZ). This zone produced three of the largest earthquakes in North America in 1811--1812. Presently, the NMSZ is still active and is considered by engineers, seismologists, and public officials as the most hazardous seismic zone in the central and eastern United States. Bridges in the Memphis area were generally not designed for seismic resistance until 1990. Therefore, the majority of existing bridges might suffer damage from earthquakes occurring in the NMSZ. The overall objective of this study is to evaluate the expected damage to bridges and roadways on the major routes in Memphis and Shelby County resulting from New Madrid earthquakes with the aid of geographic information system (GIS) technology. The road network selected for this study includes all the Interstate highway system, all the primary and secondary routes maintained by the state, and most of the major arterial routes. There are 452 bridges on the selected roadway systems and data pertinent to these bridges and roadway systems were collected and implemented as a GIS database. The bridges in the Memphis area were classified into several types and damage states were determined

France's seismic zoning

International Nuclear Information System (INIS)

Mohammadioun, B.

1997-01-01

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

Seismic Behavior and Retrofit of Concrete Columns of Old R.C. Buildings Reinforced With Plain Bars

International Nuclear Information System (INIS)

Marefat, M. S.; Arani, K. Karbasi; Shirazi, S. M. Hassanzadeh; Amrollahi, A.

2008-01-01

Seismic rehabilitation of old buildings has been a major challenge in recent years. The first step in seismic rehabilitation is evaluation of the existing capacity and the seismic behaviour. For investigation of the seismic behaviour of RC members of a real old building in Iran which has been designed and constructed by European engineers in 1940, three half-scale column specimens reinforced with plain bars have been tested. The tests indicate significant differences between the responses of specimens reinforced by plain bars relative to those reinforced by deformed bars. A regular pattern of cracking and a relatively brittle behaviour was observed while a relatively large residual strength appeared after sudden drop of initial strength and stiffness due to slip of longitudinal bars

Using Evaluation for CBNRM Capacity Development (Southeast Asia)

International Development Research Centre (IDRC) Digital Library (Canada)

Using Evaluation for CBNRM Capacity Development (Southeast Asia) ... for evaluating both the process and outcome of capacity development efforts in CBNRM. ... Cluster Case Studies Planning Workshop, Beijing, September 21-23, 2006; ...

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

International Nuclear Information System (INIS)

Xu, J.; Graves, H.

1990-01-01

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

Groundwater environmental capacity and its evaluation index.

Science.gov (United States)

Xing, Li Ting; Wu, Qiang; Ye, Chun He; Ye, Nan

2010-10-01

To date, no unified and acknowledged definition or well-developed evaluation index system of groundwater environment capacity can be found in the academia at home or abroad. The article explores the meaning of water environment capacity, and analyzes the environmental effects caused by the exploitation of groundwater resources. This research defines groundwater environmental capacity as a critical value in terms of time and space, according to which the groundwater system responds to the external influences within certain goal constraint. On the basis of observing the principles of being scientific, dominant, measurable, and applicable, six level 1 evaluation indexes and 11 constraint factors are established. Taking Jinan spring region for a case study, this research will adopt groundwater level and spring flow as constraint factors, and the allowable groundwater yield as the critical value of groundwater environmental capacity, prove the dynamic changeability and its indicating function of groundwater environmental capacity through calculation, and finally point out the development trends of researches on groundwater environmental capacity.

Modification of evaluation response spectrum by ductility of equipment anchorage

International Nuclear Information System (INIS)

Choi, I. G.; Jun, Y. S.; Su, J. M.

2003-01-01

The failure mode of welded anchorage is assumed as brittle in the seismic capacity evaluation of nuclear power plant equipments. But the welded anchorage has some ductile capacity. This limited displacement capacity can cause the reduction of the effective frequency of high frequency equipments and the increase of the inelastic energy absorption capacity due to the nonlinear behavior. In this study, the uniform hazard spectrum for Korean nuclear power plant site was modified using the response spectrum reduction factor developed by EPRI. The spectral acceleration for various damping ratio was determined by the theoretical method based on the random vibration theory. In conclusion, the high frequency components of evaluation response spectra were greatly reduced due to the consideration of welded anchorage ductility. This reduced response spectra can be used for the development of in-structure response spectra used in the seismic capacity evaluation of high frequency equipments

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Science.gov (United States)

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

2017-11-01

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

Seismic Evaluation of Structural Insulated Panels in Comparison with Wood-Frame Panels

Directory of Open Access Journals (Sweden)

Stefanie Terentiuk

2014-07-01

Full Text Available Structural Insulated Panel (SIP wall systems have been used in residential and light commercial buildings for the past sixty years. Lack of sufficient published research on racking load performance and limited understanding of the influence of fastener types on seismic response has been a deterrent in widespread use of the wall system in seismically active areas. This paper presents the results of a study involving a total of twenty one 2.4 m × 2.4 m shear walls tested under monotonic and cyclic loading. Four different 114 mm thick SIP panel configurations and one traditional wood frame wall were tested under monotonic loading according to ASTM E 564-06; and thirteen 114 mm thick SIP panels and three wood frame walls were tested under the CUREE loading protocol according to ASTM E 2126-11. Parameters such as fastener type; spline design; hold-down anchor location; and sheathing bearing were adjusted throughout the testing in order to determine their effects on the SIP’s performance. Performance parameters such as peak load and displacement; energy dissipation; allowable drift load capacity and seismic compatibility were determined for all of the specimens. Such parameters were then used to demonstrate the SIP walls’ compatibility with the wood frame walls and to determine the efficiency of the different SIP wall configuration and spline systems employed.

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

International Nuclear Information System (INIS)

Sugino, Hideharu; Onizawa, Kunio; Suzuki, Masahide

2005-09-01

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

Seismic risk evaluation within the technology neutral framework

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Science.gov (United States)

2008-06-01

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

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

Relevance of normative values for functional capacity evaluation

NARCIS (Netherlands)

Soer, R.; Van Der Schans, C.; Geertzen, J.; Groothoff, J.; Brouwer, Sandra; Dijkstra, P.; Reneman, M.

2009-01-01

Background: Functional Capacity Evaluations (FCEs) are evaluations designed to measure capacity to perform activities and are used to make recommendations for participation in work. Normative values of healthy working subjects' performances are unavailable, thus patients' performances cannot be

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

International Nuclear Information System (INIS)

SWENSON, C.E.

2000-01-01

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

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

Science.gov (United States)

2011-09-16

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

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

International Nuclear Information System (INIS)

1996-01-01

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

Mentoring approach improves evaluation capacity of ICTD ...

International Development Research Centre (IDRC) Digital Library (Canada)

2016-06-09

Jun 9, 2016 ... ... initiative is improving evaluation capacities of researchers studying Information and ... Capacity in ICTD (DECI) provides researchers from five IDRC-funded projects in ... Using technology to deliver quality education in Asia.

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

Science.gov (United States)

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

2015-04-01

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

Review of Seismic Evaluation Methodologies for Nuclear Power Plants Based on a Benchmark Exercise

International Nuclear Information System (INIS)

2013-11-01

Niigataken-chuetsu-oki (NCO) earthquake (Mw = 6.6) occurred on 16 July 2007 and affected the Kashiwazaki-Kariwa (K-K) NPP in Japan. Although there was significant loss of main shock data due to transmission problems, a significant number of instruments were still able to measure the acceleration at different locations in soil (boreholes) and in structures at the K-K NPP during the main shock and the aftershocks. The availability of all these instrumental data provided an excellent background for initiating a benchmarking exercise known as the KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment (KARISMA). The main objective of the KARISMA benchmark exercise is to study a comparison between analytical seismic response versus real response of selected structure, system and components (SSCs) of K-K NPP Unit 7. The KARISMA benchmark exercise includes benchmarking the analytical tools and numerical simulation techniques used for predicting seismic response of NPP structures (in linear and non-linear ranges), site response, soil-structure interaction phenomena, seismic response of piping systems, 'sloshing' in the spent fuel pool and buckling of tanks. The benchmark is primarily based on data provided by Tokyo Electric Power Company (TEPCO). It is not linked to the seismic re-evaluation of K-K NPP carried out by TEPCO. Twenty-one organizations, comprising researchers, operating organizations, regulatory authorities, vendors and technical support organizations from 14 countries, participated in the benchmarking exercises. This publication, including a CD-ROM, summarizes the analyses of the main results of the benchmarking exercise for the K-K NPP reactor building (including static and modal analyses of the fixed base model, soil column analyses, analyses of the soil-structure models and margin assessment of the K-K NPP reactor building), the analyses of the main results of the benchmarking exercise for the residual heat removal piping system (including

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

Directory of Open Access Journals (Sweden)

Dongxu Hou

2014-01-01

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

Relay testing parametric investigation of seismic fragility

International Nuclear Information System (INIS)

Bandyopadhyay, K.; Hofmayer, C.; Kassir, M.; Pepper, S.

1989-01-01

The seismic capacity of most electrical equipment is governed by malfunction of relays. An evaluation of the existing relay test data base at Brookhaven National Laboratory (BNL) has indicated that the seismic fragility of a relay may depend on various parameters related to the design or the input motion. In particular, the electrical mode, contact state, adjustment, chatter duration acceptance limit, and the frequency and the direction of the vibration input have been considered to influence the relay fragility level. For a particular relay type, the dynamics of its moving parts depends on the exact model number and vintage and hence, these parameters may also influence the fragility level. In order to investigate the effect of most of these parameters on the seismic fragility level, BNL has conducted a relay test program. The testing has been performed at Wyle Laboratories. Establishing the correlation between the single frequency fragility test input and the corresponding multifrequency response spectrum (TRS) is also an objective of this test program. This paper discusses the methodology used for testing and presents a brief summary of important test results. 1 ref., 10 figs

Seismic evaluation of vulnerability for SAMA educational buildings in Tehran

International Nuclear Information System (INIS)

Amini, Omid Nassiri; Amiri, Javad Vaseghi

2008-01-01

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

Promoting evaluation capacity building in a complex adaptive system.

Science.gov (United States)

Lawrenz, Frances; Kollmann, Elizabeth Kunz; King, Jean A; Bequette, Marjorie; Pattison, Scott; Nelson, Amy Grack; Cohn, Sarah; Cardiel, Christopher L B; Iacovelli, Stephanie; Eliou, Gayra Ostgaard; Goss, Juli; Causey, Lauren; Sinkey, Anne; Beyer, Marta; Francisco, Melanie

2018-04-10

This study provides results from an NSF funded, four year, case study about evaluation capacity building in a complex adaptive system, the Nanoscale Informal Science Education Network (NISE Net). The results of the Complex Adaptive Systems as a Model for Network Evaluations (CASNET) project indicate that complex adaptive system concepts help to explain evaluation capacity building in a network. The NISE Network was found to be a complex learning system that was supportive of evaluation capacity building through feedback loops that provided for information sharing and interaction. Participants in the system had different levels of and sources of evaluation knowledge. To be successful at building capacity, the system needed to have a balance between both centralized and decentralized control, coherence, redundancy, and diversity. Embeddedness of individuals within the system also provided support and moved the capacity of the system forward. Finally, success depended on attention being paid to the control of resources. Implications of these findings are discussed. Copyright © 2018 Elsevier Ltd. All rights reserved.

Study on Seismic Behavior of Recycled Concrete Energy-efficient Homes Structure Wall

Directory of Open Access Journals (Sweden)

Dong Lan

2016-01-01

Full Text Available The main point is to study the seismic behavior of the lattice type recycled concrete energy saving wall under low-cyclic loading,to provide the basis for the seismic performance of application of recycled concrete lattice wall in energy-saving residential structure. Design two walls with the same structure measures, include Lattice type recycled concrete wall and natural concrete wall, they are tested under low-cycle repetitive loading, compared failure mode and seismic performance in different reinforcement conditions of side column. The bearing capacity and ductility of recycled aggregate concrete are better than natural aggregate concrete, The stiffness degradation curves and the skeleton curves of the walls are basically the same, both of them have better seismic energy dissipation capacity. Lattice type concrete wall is good at seismic performance, recycled aggregate concrete is good at plastic deformation ability, it is advantageous to seismic energy dissipation of wall, it can be applied in energy efficient residential structure wall.

Development of seismic hazard analysis in Japan

International Nuclear Information System (INIS)

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

1987-01-01

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

Ultimate load capacities of expansion anchor bolts

International Nuclear Information System (INIS)

Czarnecki, R.M.; Manrique, M.A.; Samaddar, S.K.

1993-01-01

A summary of available experimental expansion anchor bolt test data is presented. These data were collected as part of programs by the nuclear industry to address generic issues related to verification of seismic adequacy of equipment in nuclear power plants. Some of the data presented are suitable for use in seismic probabilistic risk assessments. For example, mean values of ultimate strength, along with their standard deviation and coefficients of variation, for a range of most typical expansion anchor bolt sizes are presented. Effects of interaction between shear and tension, edge distance, spacing, and cracking of the concrete are presented in a manner that is more suitable for use in deterministic evaluations. Related industry programs to derive anchor bolt capacities are briefly discussed. Recommendations for areas of further investigation are also presented

Seismic fragility of a reinforced concrete structure

Energy Technology Data Exchange (ETDEWEB)

Kurmann, Davide [Axpo Power AG, Baden (Switzerland); Proske, Dirk [Axpo Power AG, Doettingen (Switzerland); Cervenka, Jan [Cervenka Consulting, Prague (Czech Republic)

2013-05-15

Structures can be exposed to seismic loading. For structures of major importance, extreme seismic loadings have to be considered. The proof of safety for such loadings requires sophisticated analysis. This paper introduces an analysis method which of course still includes simplifications, but yields to a far more realistic estimation of the seismic load bearing capacity of reinforced concrete structures compared to common methods. It is based on the development of pushover curves and the application of time-histories for the dynamic model to a representative harmonic oscillator. Dynamic parameters of the oscillator, such as modal mass and damping are computed using a soil-structure-interaction analysis. Based on the pushover-curve nonlinear force-deformation-capacities are applied to the oscillator including hysteresis behaviour characteristics. The oscillator is then exposed to time-histories of several earthquakes. Based on this computation the ductility is computed. The ductility can be scaled based upon the scaling of the time-histories. Since both, the uncertainty of the earthquake by using different timehistories and the uncertainty of the structure by using characteristic and mean material values, are considered, the uncertainty of the structure under seismic loading can be explicitly represented by a fragility. (orig.)

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

Multi-lane Roundabout Capacity Evaluation

Directory of Open Access Journals (Sweden)

Ammar Šarić

2017-07-01

Full Text Available Although two-lane roundabouts theoretically exhibit excellent operating performance, in practice, safety problems arise because of inappropriate driving behavior. Turbo roundabouts, which are characterized by a much higher level of safety, are alternatives to classic two-lane roundabouts, but the capacity-related benefits derived from such roundabouts remain an open issue. Accordingly, this study uses an equilibrium traffic flow allocation approach to evaluate multi-lane roundabout capacity based on gap acceptance theory. Capacity levels are calculated and compared for different gap acceptance parameters, including local parameters, and different traffic flow scenarios. It is found that the capacity of minor approaches on turbo roundabouts is always higher than on two-lane roundabouts, but that the main approaches on two-lane roundabouts exhibit better performance in terms of fully equilibrium traffic allocation. This state, however, cannot be achieved for every demand scenario. The results depend strongly on traffic movements and gap acceptance parameters indicating the need for local calibration processes.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Screening dynamic evaluation of SRS cooling water line

International Nuclear Information System (INIS)

Bezler, P.; Shteyngart, S.; Breidenbach, G.

1991-01-01

The production reactors at the Savannah River Site (SRS) have been shut down due to perceived safety concerns. A major concern is the seismic integrity of the plant. A comprehensive program is underway to assess the seismic capacity of the existing systems and components and to upgrade them to acceptable levels. The evaluation of the piping systems at the SRS is a major element of this program. Many of the piping systems at the production reactors were designed without performing dynamic analyses. Instead their design complied with good design practice for dead weight supported systems with proper accommodation of thermal expansion effects. In order to gain some insight as to the seismic capacity of piping installed in this fashion, dynamic analyses were performed for some lines. Since the piping was not seismically supported, the evaluations involved various approximations and the results are only used as a screening test of seismic adequacy. In this paper, the screening evaluations performed for the raw water inlet line are described. This line was selected for evaluation since it was considered typical of the smaller diameter piping systems at the plant. It is a dead weight supported system made up of a run of small diameter piping which extends for great distances over many dead weight supports and through wall penetrations. The results of several evaluations for the system using different approximations to represent the support system are described. 2 figs., 4 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-04-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Using Evaluation for CBNRM Capacity Development (Southeast Asia)

International Development Research Centre (IDRC) Digital Library (Canada)

Using Evaluation for CBNRM Capacity Development (Southeast Asia). Recently, capacity development has regained a central place on the agenda of many donor organizations, including IDRC. The challenge is to measure the results of capacity development initiatives. In Asia, there are a few ongoing efforts to document ...

Seismic enhancement of multi-span continuous bridges subjected to three-directional excitations

Science.gov (United States)

Aryan, H.; Ghassemieh, M.

2015-04-01

Considering the seismic ground motions as the excitations in only two principal horizontal directions of the bridges and ignoring the third vertical direction is a disregard for the seismic conditions of the region and the bridge distance from epicenter. Numerous cases of substantial damages have been reported among the bridges tremendously suffered from being exposed to the simultaneous three-directional seismic ground motions. Besides the significant compression and tension damages in the columns due to the presence of vertical excitation, it could lead to unexpected shear and flexural failures in the columns and other components as well. Because the axial force variation in the columns due to three-directional excitations, could affect the demands and capacities of the bridge’s components. With respect to this issue, several studies on the bridge damages during the earthquakes have urged researchers to offer efficient methods for bridges handling of the three-directional seismic excitations. Thus, this paper presents and evaluates a superelastic based system for designing as well as retrofitting the multi-span continuous (MSC) bridges that can cope with two- and three-directional seismic excitations. Efficiency evaluation of the proposed system is conducted through various nonlinear time history analyses on a three-dimensional model of a detailed MSC bridge using a suite of developed ground motions for the bridge region. Also, all the analyses are fulfilled based on variation of one influential design characteristic of the proposed system in order to achieve the optimal design. Several pertinent assessment parameters are used during the evaluation of the proposed system. Finally, the efficiency of the new system subjected to the vertical and horizontal seismic excitations is confirmed according to reduction of the bridge responses and improvement in nonlinear performance of the columns in comparison with the as-built bridge results.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Evaluation of seismic stability of near field

International Nuclear Information System (INIS)

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

1999-11-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Research items regarding seismic residual risk evaluation

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Some Underexamined Aspects of Evaluation Capacity Building

Science.gov (United States)

Leviton, Laura C.

2014-01-01

Evaluation capacity building (ECB) has progressed as a concept since it was the conference theme of the American Evaluation Association in the year 2000. This commentary poses some questions about underexamined issues in ECB about organizations, evaluators, and funders.

The Ultimate Pile Bearing Capacity from Conventional and Spectral Analysis of Surface Wave (SASW) Measurements

Science.gov (United States)

Faizah Bawadi, Nor; Anuar, Shamilah; Rahim, Mustaqqim A.; Mansor, A. Faizal

2018-03-01

A conventional and seismic method for determining the ultimate pile bearing capacity was proposed and compared. The Spectral Analysis of Surface Wave (SASW) method is one of the non-destructive seismic techniques that do not require drilling and sampling of soils, was used in the determination of shear wave velocity (Vs) and damping (D) profile of soil. The soil strength was found to be directly proportional to the Vs and its value has been successfully applied to obtain shallow bearing capacity empirically. A method is proposed in this study to determine the pile bearing capacity using Vs and D measurements for the design of pile and also as an alternative method to verify the bearing capacity from the other conventional methods of evaluation. The objectives of this study are to determine Vs and D profile through frequency response data from SASW measurements and to compare pile bearing capacities obtained from the method carried out and conventional methods. All SASW test arrays were conducted near the borehole and location of conventional pile load tests. In obtaining skin and end bearing pile resistance, the Hardin and Drnevich equation has been used with reference strains obtained from the method proposed by Abbiss. Back analysis results of pile bearing capacities from SASW were found to be 18981 kN and 4947 kN compared to 18014 kN and 4633 kN of IPLT with differences of 5% and 6% for Damansara and Kuala Lumpur test sites, respectively. The results of this study indicate that the seismic method proposed in this study has the potential to be used in estimating the pile bearing capacity.

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Roglans, J.; Wang, C.Y.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Seismic assessment and upgrading of nuclear power plants in Eastern Europe

Energy Technology Data Exchange (ETDEWEB)

Katona, T; Kostov, M

1997-03-01

The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

Seismic assessment and upgrading of nuclear power plants in Eastern Europe

International Nuclear Information System (INIS)

Katona, T.; Kostov, M.

1997-01-01

The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

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

Science.gov (United States)

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

2018-05-01

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

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

International Nuclear Information System (INIS)

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

1990-07-01

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

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

Evaluation of Floor Response Spectrum considering Ductility of Structure

Energy Technology Data Exchange (ETDEWEB)

Park, Junhee; Choi, In-Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The FRS (floor response spectrum) is directly influenced by the behavior of structure under the seismic load. If the structure is nonlinear range, the energy dissipation will be occurred by the damage of structure and the maximum force will be reduced. In Zion method, the inelastic energy abortion factor has been used to consider the nonlinearity of structure. This factor was used for the seismic fragility of structure. For the seismic fragility of equipment, the uncertainty of this factor was used differently according to the story level. But this method is not warranted under the strong earthquake leads to the structural damage. Therefore it is needed to evaluate the FRS considering the nonlinear behavior of structure and to assessment the conservatism related to nonlinear behavior of structure in FRS. In this study, the nonlinear analysis was performed for the conservatism of FRS under the damage of structure. The conservatism of FRS by the nonlinear analysis was compared by that proposed by the Zion method. The conservatism of floor acceleration response for the equipment was evaluated by performing the nonlinear analysis. From the nonlinear analysis results, it was showed that the median and β{sub c} of FRSR was increased with the ductility of structure and the response of equipment had the resonance effect between the frequency of equipment and structure. The seismic capacity of equipment by the Zion method can be different from the real seismic capacity of equipment because the inelastic structure response factor has nothing to do with the ductility of structure. Therefore the median and COV for FRSR should be defined considering the ductility of structure and the frequency of equipment for more exactly evaluating the seismic capacity of equipment.

Evaluation of Floor Response Spectrum considering Ductility of Structure

International Nuclear Information System (INIS)

Park, Junhee; Choi, In-Kil

2014-01-01

The FRS (floor response spectrum) is directly influenced by the behavior of structure under the seismic load. If the structure is nonlinear range, the energy dissipation will be occurred by the damage of structure and the maximum force will be reduced. In Zion method, the inelastic energy abortion factor has been used to consider the nonlinearity of structure. This factor was used for the seismic fragility of structure. For the seismic fragility of equipment, the uncertainty of this factor was used differently according to the story level. But this method is not warranted under the strong earthquake leads to the structural damage. Therefore it is needed to evaluate the FRS considering the nonlinear behavior of structure and to assessment the conservatism related to nonlinear behavior of structure in FRS. In this study, the nonlinear analysis was performed for the conservatism of FRS under the damage of structure. The conservatism of FRS by the nonlinear analysis was compared by that proposed by the Zion method. The conservatism of floor acceleration response for the equipment was evaluated by performing the nonlinear analysis. From the nonlinear analysis results, it was showed that the median and β c of FRSR was increased with the ductility of structure and the response of equipment had the resonance effect between the frequency of equipment and structure. The seismic capacity of equipment by the Zion method can be different from the real seismic capacity of equipment because the inelastic structure response factor has nothing to do with the ductility of structure. Therefore the median and COV for FRSR should be defined considering the ductility of structure and the frequency of equipment for more exactly evaluating the seismic capacity of equipment

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

Science.gov (United States)

Nishikawa, Hayato; Miyajima, Masakatsu

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1996-03-01

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

Conceptual design by analysis of KALIMER seismic isolation

International Nuclear Information System (INIS)

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

1996-06-01

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

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

Directory of Open Access Journals (Sweden)

Naik Muhammad

2017-10-01

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

Simplified static method for determining seismic loads on equipment in moderate and high hazard facilities

International Nuclear Information System (INIS)

Scott, M.A.; Holmes, P.A.

1991-01-01

A simplified static analysis methodology is presented for qualifying equipment in moderate and high-hazard facility-use category structures, where the facility use is defined in Design and Evaluation Guidelines for Department of Energy Facilities Subjected to Natural Phenomena Hazards, UCRL-15910. Currently there are no equivalent simplified static methods for determining seismic loads on equipment in these facility use categories without completing dynamic analysis of the facility to obtain local floor accelerations or spectra. The requirements of UCRL-15910 specify the use of open-quotes dynamicclose quotes analysis methods, consistent with Seismic Design Guidelines for Essential Buildings, Chapter 6, open-quotes Nonstructural Elements,close quotes TM5-809-10-1, be used for determining seismic loads on mechanical equipment and components. Chapter 6 assumes that the dynamic analysis of the facility has generated either floor response spectra or model floor accelerations. These in turn are utilized with the dynamic modification factor and the actual demand and capacity ratios to determine equipment loading. This complex methodology may be necessary to determine more exacting loads for hard to qualify equipment but does not provide a simple conservative loading methodology for equipment with ample structural capacity

HANFORD DOUBLE SHELL TANK THERMAL AND SEISMIC PROJECT SUMMARY OF COMBINED THERMAL AND OPERATING LOADS WITH SEISMIC ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

MACKEY TC; DEIBLER JE; RINKER MW; JOHNSON KI; ABATT FG; KARRI NK; PILLI SP; STOOPS KL

2009-01-15

design waste temperature of 350 F and the full 60-year corrosion allowance on the tank wall of 0.060 inch. However, analysis at a more realistic temperature of 250 F or corrosion allowance of 0.025 inch results in an acceptable demand/capacity ratio according to the ASME code criteria. Thus, buckling of the primary tank is judged to be unlikely for the current lack of corrosion in the tanks, and the expectation that the maximum waste temperature will not exceed 210 F. The reinforced concrete structure was evaluated as specified by the American Concrete Institute (ACI) code requirements for nuclear safety-related structures (ACI-349). The demand was demonstrated to be lower than the capacity at all locations. Revision 1 is being issued to document changes to the anchor bolt evaluation. RPP-RPT-32237 Rev. 1, Hanford Double-Shell Tank Thermal and Seismic Project-Increased Liquid Level Analysis for 241AP Tank Farms, described changes to the anchor bolt modeling and evaluation which were implemented in response to the independent reviewer's comments. Similar changes have been made in the bounding tank analysis and are documented in RPP-RPT-28968 Rev. 1. The conclusions of the previous releases of this report remain unchanged.

The influence of backfill on seismicity

CSIR Research Space (South Africa)

Hemp, DA

1990-09-01

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

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

New Evaluation of Seismic Hazard in Cental America and la Hispaniola

Science.gov (United States)

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

2013-05-01

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

Modelling of Rotational Capacity in Reinforced Linear Elements

DEFF Research Database (Denmark)

Hestbech, Lars; Hagsten, Lars German; Fisker, Jakob

2011-01-01

on the rotational capacity of the plastic hinges. The documentation of ductility can be a difficult task as modelling of rotational capacity in plastic hinges of frames is not fully developed. On the basis of the Theory of Plasticity a model is developed to determine rotational capacity in plastic hinges in linear......The Capacity Design Method forms the basis of several seismic design codes. This design philosophy allows plastic deformations in order to decrease seismic demands in structures. However, these plastic deformations must be localized in certain zones where ductility requirements can be documented...... reinforced concrete elements. The model is taking several important parameters into account. Empirical values is avoided which is considered an advantage compared to previous models. Furthermore, the model includes force variations in the reinforcement due to moment distributions and shear as well...

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

Science.gov (United States)

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

2017-06-01

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

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

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

CERN Document Server

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

2002-01-01

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

SHAM: High-level seismic tests of piping at the HDR

International Nuclear Information System (INIS)

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.; Malcher, L.; Schrammel, D.; Steinhilber, H.; Costello, J.F.

1988-01-01

As part of the second phase of vibrational/earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, high-level simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) in-plant piping system with two servohydraulic actuators, each capable of generating 40 tons of force. The purpose of these experiments was to study the behavior of piping subjected to seismic excitation levels that exceed design levels manifold and may result in failure/plastification of pipe supports and pipe elements, and to establish seismic margins for piping and pipe supports. The performance of six different dynamic pipe support systems was compared in these tests and the response, operability, and fragility of dynamic supports and of a typical US gate valve were investigated. Data obtained in the tests are used to validate analysis methods. Very preliminary evaluations lead to the observation that, in general, failures of dynamic supports (in particular snubbers) occur only at load levels that substantially exceed the design capacity. Pipe strains at load levels exceeding the design level threefold are quite small, and even when exceeding the design level eightfold are quite tolerable. Hence, under seismic loading, even at extreme levels and in spite of multiple support failures, pipe failure is unlikely. 5 refs., 16 figs

Seismic microzonation of Bangalore, India

Indian Academy of Sciences (India)

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

Understanding and Measuring Evaluation Capacity: A Model and Instrument Validation Study

Science.gov (United States)

Taylor-Ritzler, Tina; Suarez-Balcazar, Yolanda; Garcia-Iriarte, Edurne; Henry, David B.; Balcazar, Fabricio E.

2013-01-01

This study describes the development and validation of the Evaluation Capacity Assessment Instrument (ECAI), a measure designed to assess evaluation capacity among staff of nonprofit organizations that is based on a synthesis model of evaluation capacity. One hundred and sixty-nine staff of nonprofit organizations completed the ECAI. The 68-item…

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

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

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

International Nuclear Information System (INIS)

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

1990-07-01

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

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

Science.gov (United States)

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

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

Organizational Capacity to Do and Use Evaluation: Results of a Pan-Canadian Survey of Evaluators

Science.gov (United States)

Cousins, J. Bradley; Elliott, Catherine; Amo, Courtney; Bourgeois, Isabelle; Chouinard, Jill; Goh, Swee C.; Lahey, Robert

2008-01-01

Despite increasing interest in the integration of evaluative inquiry into organizational functions and culture, the availability of empirical research addressing organizational capacity building to do and use evaluation is limited. This exploratory descriptive survey of internal evaluators in Canada asked about evaluation capacity building in the…

Seismic evaluation of critical facilities at the Lawrence Livermore Laboratory

International Nuclear Information System (INIS)

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

1976-01-01

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

Seismic evaluation of nuclear installations; Avaliacao sismica de instalacoes nucleares

Energy Technology Data Exchange (ETDEWEB)

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

1997-10-01

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

Seismic Isolation Working Meeting Gap Analysis Report

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-01

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

Seismic evaluation of existing nuclear power plants and other facilities V. 1. Proceedings of the technical committee meeting. Working material

International Nuclear Information System (INIS)

2002-01-01

The objectives of this Tcm are: to review the IAEA Safety Report on Seismic Evaluation of Existing Nuclear Power Plants in order to achieve a consensus among Member States on this matter and to discuss the outlines of an IAEA Co-ordinated Research Programme on specific topics related to this subject. Today the nuclear industry relies much more on existing facilities than on the design of new ones. Nevertheless it appears that safety evaluation against external hazards is not a decreasing activity. The reason being that maintaining an acceptable level of nuclear safety requires periodic re-assessments of facilities, either because of modifications of the environment due to human activity or because of new data or new approaches in the assessment of natural hazards. In this regard, seismic re-evaluation has increasingly become a key issue for several existing nuclear facilities, including not only nuclear power plants but also other plants of the fuel cycle, as well as research reactors or laboratories. The IAEA has already supported development of engineering practices in this field by managing a Co-ordinated Research Programme, launched in 1992, on a Benchmark Study for the seismic analysis and testing of WWER 1000 type NPPs. It is now proposed to investigate other aspects of this issue. Many of these facilities were built according to older standards which did not take into account seismic hazard. Consequently, the seismic re-evaluation of existing facilities is a real challenge for earthquake engineers. In most of the cases, it is impossible to re-evaluate according to the up to date standards because entering these standards implies that some design rules are met, what is generally not the case for older facilities. In the best cases some rules exist for non nuclear buildings. In order to achieve a consensus on this matter, the IAEA intends to edit a Safety Report on 'Seismic Evaluation of existing NPPs'. The TCM will offer the opportunity to review the draft of

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)

A report on seismic re-evaluation of Cirus systems

International Nuclear Information System (INIS)

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

2003-06-01

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

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

International Nuclear Information System (INIS)

1991-01-01

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

Seismic component fragility data base for IPEEE

International Nuclear Information System (INIS)

Bandyopadhyay, K.; Hofmayer, C.

1990-01-01

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

Seismic qualification of multiple interconnected safety-related cabinets in a high seismic zone

International Nuclear Information System (INIS)

Khan, M.R.; Chen, W.H.W.; Wang, T.Y.

1993-01-01

Certain safety-related multiple, interconnected electrical cabinets and the devices contained therein are required to perform their intended safety functions during and after a design basis seismic event. In general, seismic testing is performed to ensure the structural integrity of the cabinets and the functionality of their associated devices. Constrained by the shake table capacity, seismic testing is usually performed only for a limited number of interconnected cabinets. Also, original shake table tests performed usually did not provide detailed response information at various locations inside the cabinets. For operational and maintenance purposes, doors and panels of some cabinets may need to be opened while the adjacent cabinets are required to remain functional. In addition, in-cabinet response spectra need to be generated for the seismic qualification of new devices and the replacement parts. Consequently, seismic analysis of safety-related multiple, interconnected cabinets is frequently required for configurations which are different from the original tested conditions. This paper presents results of seismic tests of three interconnected safety-related cabinets and finite element analyses performed to compare the analytical results with those obtained from the cabinet seismic tests. Parametric analyses are performed to determine how many panels and doors can be opened while the adjacent cabinets still remain functional. The study indicates that for cabinets located in a high seismic zone, the critical damping of the cabinet is significantly higher than 5% to 7% typically used in qualifying electrical equipment. For devices mounted on the cabinet doors to performed their intended safety function, it requires stiffening of doors and that these doors be properly bolted to the cabinet frame. It also shows that even though doors and panels bolted to the cabinet frame are the primary seismic resistant element of the cabinet, opening of a limited number of them

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

Energy Technology Data Exchange (ETDEWEB)

Berge-Thierry, C

2007-05-15

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

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

Seismic (SSE) evaluation for the 291Z stack at the Hanford Site -- Addition of environmental monitoring penetrations

International Nuclear Information System (INIS)

Baxter, J.T.

1994-01-01

The purpose of this 291Z stack analysis is to determine the structural effects of chipping additional holes into the stacks concrete walls. The proposed holes are for new environmental monitoring sample probes to be installed at three different elevations. The approximate elevations proposed at this time are 50 ft, 135 ft and 175 ft. There will be four holes required at each of the elevations to support two sample probes extending across the diameter of the stack. A structural sensitivity study has been completed to assess the effect of the proposed holes on the baseline seismic qualification of the stack completed by URS/John A. Blume ampersand Associates, Engineers, San Francisco, California (URS/Blume) in August, 1988. Results of the sensitivity study indicate that the stack would still have adequate structural moment capacity if the new holes were drilled cutting the vertical strength reinforcing steel, or if existing penetrations added since original construction have inadvertently cut vertical rebars. For current and future modifications, no vertical rebar should be cut. A limited number of horizontal rebar, no more than 2, may be cut at the new hole locations without significantly influencing the stack structural shear capacity. New penetrations in the 291Z stack should not be located below elevation 47 ft., 4 in. due to rebar layout and the fact that maximum seismic structural loads occur below this elevation. No vertical rebar should be cut when chipping the new penetrations in the stack concrete wall for the environmental monitoring equipment. Wind load qualification was reviewed. Seismic loads govern over wind loads for all structural load cases; therefore no additional wind analyses are required

[Assessment of Functioning when Conducting Occupational Capacity Evaluations--What is "Evidence-Based"?].

Science.gov (United States)

Canela, Carlos; Schleifer, Roman; Dube, Anish; Hengartner, Michael P; Ebner, Gerhard; Seifritz, Erich; Liebrenz, Michael

2016-03-01

Occupational capacity evaluations have previously been subject to criticism for lacking in quality and consistency. To the authors' knowledge, there is no clear consensus on the best way to formally assess functioning within capacity evaluations. In this review we investigated different instruments that are used to assess functioning in occupational capacity evaluations. Systematic review of the literature. Though several instruments that assess functional capacity were found in our search, a specific validated instrument assessing occupational capacity as part of a larger psychiatric evaluation was not found. The limitations of the existing instruments on assessing functional capacity are discussed. Medical experts relying on instruments to conduct functional capacity evaluations should be cognizant of their limitations. The findings call for the development and use of an instrument specifically designed to assess the functional and occupational capacity of psychiatric patients, which is also likely to improve the quality of these reports. © Georg Thieme Verlag KG Stuttgart · New York.

Capacity value evaluation of photovoltaic power generation

International Nuclear Information System (INIS)

Kurihara, I.

1993-01-01

The paper presents an example of capacity value (kW-value) evaluation of photovoltaic generation from power companies generation planning point of view. The method actually applied to evaluate the supplying capability of conventional generation plants is briefly described. 21 figs, 1 tab

Evaluation of the seismic integrity of a plutonium-handling facility

International Nuclear Information System (INIS)

Coats, D.W.

1981-01-01

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

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

International Nuclear Information System (INIS)

Berge-Thierry, C.

2007-05-01

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

Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

International Nuclear Information System (INIS)

Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

1995-01-01

This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

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

International Nuclear Information System (INIS)

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

1990-07-01

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

Seismic safety of Paks nuclear power plant

International Nuclear Information System (INIS)

Katona, T.

1993-01-01

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

Seismic performance evaluation of multi-span existing masonry arch bridge

Science.gov (United States)

Laterza, Michelangelo; D'Amato, Michele; Casamassima, Vito Michele

2017-07-01

Existing old masonry arch bridges represent an architectural and cultural heritage of inestimable value, assuming nowadays an important strategic role since most of them are still in service and link roads of primary importance for vehicular traffic. They were mostly built in the last century without considering any horizontal action, and nowadays are serving roads characterized by a transit loads certainly heavier and more frequent than the ones of past. Moreover, very often due to absence of maintenance and to weathering conditions, the elements deteriorate more and more with a consequent loss of integrity and reduction of their carrying capacity. In this paper the seismic assessment of an old multi span masonry arch bridge still in service is illustrated. Pushover analyses are performed with the aim to investigate the numerical model sensitivity and the influence on the global nonlinear response of the bridge components.

Seismic fragility analyses

International Nuclear Information System (INIS)

Kostov, Marin

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Berge-Thierry, C

2007-05-15

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

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

Mentoring approach improves evaluation capacity of ICTD ...

International Development Research Centre (IDRC) Digital Library (Canada)

Developing Evaluation Capacity in ICTD (DECI) provides researchers from five IDRC-funded projects in Asia ongoing mentorship to learn and apply the Utilization Focused Evaluation (UFE) approach to their projects. DECI demonstrates the value of mentoring as a training approach, where researchers are coached as they ...

Developing Evaluation Capacity in Information and Communication ...

International Development Research Centre (IDRC) Digital Library (Canada)

17 août 2009 ... The nuanced ways in which information and communication technologies (ICTs) interact with developing country societies necessitate thoughtful, systematic evaluation of their effects, grounded in the concepts of utility, outcome and ownership. This grant will allow Developing Evaluation Capacity in ICT4D ...

Light Water Reactor Sustainability Program Advanced Seismic Soil Structure Modeling

Energy Technology Data Exchange (ETDEWEB)

Bolisetti, Chandrakanth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-06-01

Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in some instances the current SPRA approach has 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). SPRA’s are performed by convolving the seismic hazard (this is the estimate of all likely damaging earthquakes at the site of interest) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, 2) fragility or capacity of structures, systems and components (SSC), and 3) systems analysis. Two areas where NLSSI effects may be important in SPRA calculations are, 1) when calculating in-structure response at the area of interest, and 2) calculation of seismic fragilities (current fragility calculations assume a lognormal distribution for probability of failure of components). Some important effects when using NLSSI in the SPRA calculation process include, 1) gapping and sliding, 2) inclined seismic waves coupled with gapping and sliding of foundations atop soil, 3) inclined seismic waves coupled with gapping and sliding of deeply embedded structures, 4) soil dilatancy, 5) soil liquefaction, 6) surface waves, 7) buoyancy, 8) concrete cracking and 9) seismic isolation The focus of the research task presented here-in is on implementation of NLSSI into the SPRA calculation process when calculating in-structure response at the area

Motor function and activities of daily living capacity of patients with fractures sustained during the Wenchuan earthquake.

Science.gov (United States)

Zhang, Jin-Long; He, Hong-Chen; Lin, Hai-Dan; Luo, Qing-Lu; Xia, Lu; Li, Sha-Sha; He, Cheng-Qi

2011-05-01

On the afternoon of May 12, 2008, a 8.0-magnitude earthquake hit Sichuan Province, a mountainous region in Western China, killing about 70 000 people and leaving over 18 000 missing. What about the survivors motor functions and activities of daily living (ADL) capacity, especially for fractures? We need the data to guide the rehabilitation for the seismic wounded and it's important to collect the data for the future. We study the survivors to understand the motor functions and ADL capacity of patients with fractures sustained in the Wenchuan earthquake, to provide a basis for rehabilitation and treatment. We used the Manual Muscle Testing method to evaluate muscle strength, the joint angle scale to measure joint range of motion (ROM), and the Barthel index to evaluate the activities of daily living status. SPSS 13.0 software was used to analyze the data and the results were tested using one-way analysis of variance (ANOVA). The number of seismic wounded amounted to 487; 81.1% of patients had fractures. Most of the injured had fractures in multiple regions (53.9% of all fracture patients), followed by fractures of the upper limb (34.0% of patients); cranial fractures were rare (2.3%). Totally 82.0% had restricted range of motion, 23.5% had decreased muscle force, and 72.2% of the patients had restricted activities of daily living capacities. With time the activities of daily living capacity of female increased (P 0.05). The difference between the patients' ages and ADL capacities did not reach statistical significance (P > 0.05), nor was there a significant difference between their ages and the numbers of days in hospital (P > 0.05). Fractures were the main issue in the seismic wounded, many of them had reductions in the ROM, muscle force and ADL capacities. The physicians involved in rehabilitation should pay greater attention to muscle force exercises, joint mobilization, and occupational therapy during the early phases post disaster.

Evaluating Seismic Activity in Ethiopia

African Journals Online (AJOL)

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

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

Science.gov (United States)

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

2018-04-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Kawai, Tadashi; Ishimaru, Makoto

2010-01-01

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

A Study of Seismic Capacity of Nuclear Equipment with Seismic Isolation System

Energy Technology Data Exchange (ETDEWEB)

Kim, Min Kyu; Choun, Young Sun; Choi, In Kil; Seo, Jeong Moon

2004-05-15

In this study, the base isolation systems for equipment are presented and the responses of each isolation system are investigated. As for the base isolation systems, a natural rubber bearing (NRB), a high damping rubber bearing (HDRB) and a friction pendulum system (FPS) are selected. The shaking table tests are carried out for three kinds of structural types. As input motions, artificial time histories enveloping the US NRC RG 1.60 spectrum and the probability-based scenario earthquake spectra developed for the Korean nuclear power plant site as well as a typical near-fault earthquake record are used. Uniaxial, biaxial, and triaxial excitations are conducted with PGAs of 0.05, 0.1, 0.2 and 0.25g. Acceleration responses are measured at the top of the equipment model and the floors using an accelerometer. The reduction of the seismic forces transmitted to the equipment models are determined for different isolation systems and input motions.

A Study of Seismic Capacity of Nuclear Equipment with Seismic Isolation System

International Nuclear Information System (INIS)

Kim, Min Kyu; Choun, Young Sun; Choi, In Kil; Seo, Jeong Moon

2004-05-01

In this study, the base isolation systems for equipment are presented and the responses of each isolation system are investigated. As for the base isolation systems, a natural rubber bearing (NRB), a high damping rubber bearing (HDRB) and a friction pendulum system (FPS) are selected. The shaking table tests are carried out for three kinds of structural types. As input motions, artificial time histories enveloping the US NRC RG 1.60 spectrum and the probability-based scenario earthquake spectra developed for the Korean nuclear power plant site as well as a typical near-fault earthquake record are used. Uniaxial, biaxial, and triaxial excitations are conducted with PGAs of 0.05, 0.1, 0.2 and 0.25g. Acceleration responses are measured at the top of the equipment model and the floors using an accelerometer. The reduction of the seismic forces transmitted to the equipment models are determined for different isolation systems and input motions

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

Energy Technology Data Exchange (ETDEWEB)

E.N. Lindner

2004-12-03

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

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

International Nuclear Information System (INIS)

E.N. Lindner

2004-01-01

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

Predicting the seismic performance of typical R/C healthcare facilities: emphasis on hospitals

Science.gov (United States)

Bilgin, Huseyin; Frangu, Idlir

2017-09-01

Reinforced concrete (RC) type of buildings constitutes an important part of the current building stock in earthquake prone countries such as Albania. Seismic response of structures during a severe earthquake plays a vital role in the extent of structural damage and resulting injuries and losses. In this context, this study evaluates the expected performance of a five-story RC healthcare facility, representative of common practice in Albania, designed according to older codes. The design was based on the code requirements used in this region during the mid-1980s. Non-linear static and dynamic time history analyses were conducted on the structural model using the Zeus NL computer program. The dynamic time history analysis was conducted with a set of ground motions from real earthquakes. The building responses were estimated in global levels. FEMA 356 criteria were used to predict the seismic performance of the building. The structural response measures such as capacity curve and inter-story drift under the set of ground motions and pushover analyses results were compared and detailed seismic performance assessment was done. The main aim of this study is considering the application and methodology for the earthquake performance assessment of existing buildings. The seismic performance of the structural model varied significantly under different ground motions. Results indicate that case study building exhibit inadequate seismic performance under different seismic excitations. In addition, reasons for the poor performance of the building is discussed.

Comparison of seismic isolation concepts for FBR

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

The seismic fragility analysis for multi-story steel structure in CANDU nuclear power plant

International Nuclear Information System (INIS)

Hwang, K.H.; Lee, B.S.; Kang, S-K.

1996-01-01

The Wolsong Unit 2 is a CANDU-6 type plant and is being constructed in the Wolsong site, where Design Basis Earthquake (DBE) was determined to be 0.2g. A seismic PSA for Wolsong Unit 2 is being performed as one of the conditions for the Construction Permit. One of the issues in the seismic PSA is the availability of the seismically non-qualified systems, which are located in the Turbine Building(T/B). Thus, the seismic fragility analysis for the T/B was performed to estimate the operability of the systems. The design seismic loads for the building were based on a ground response spectrum scaled down from the DBE to horizontal peak ground acceleration (pga) of 0.05g. The seismic fragility analysis for the building was performed using a factor of the safety method. It is estimated that the most critical failure is that of masonry walls and its High Confidence and Low Probability of Failure (HCLPF) capacity is 0.13g. The critical failure mode of the structure is identified to be tensile yielding failure of grip angle, and its HCLPF capacity is 0.34g. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Seismic qualification of civil engineering structures - Temelin NPP

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

2001-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

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

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Assessment of Seismic Vulnerability of Reinforced Concrete Frame buildings

Directory of Open Access Journals (Sweden)

Fatiha Cherifi

2018-01-01

Full Text Available The seismic activity remains strong in the north of Algeria since no less than 30 earthquakes per month are recorded. The large number of structures built before the introduction of the seismic standards represents a high seismic risk. Analysis of damage suffered during the last earthquakes highlighted the vulnerability of the existing structures. In this study the seismic behavior of the existing buildings in Tizi-Ouzou city, located in the north of Algeria, is investigated. To make this assessment, a database was created following a building inventory based on a set of technical folders and field visits. The listed buildings have been classified into different typologies. Only reinforced concrete frame buildings are considered in this paper. The approach adopted to estimate structures damage is based on four main steps: 1 construction of capacity curves using static nonlinear method “push-over”, 2 estimate of seismic hazard, 3 determination of performance points, and finally 4 deduction of damage levels.

Dynamic response of Belene WWER-1000 to seismic loading conditions

International Nuclear Information System (INIS)

Krutzik, N.J.; Petrovski, D.; Sachanski, S.

1993-01-01

Within the framework of investigating of the capacity of the WWER-100 at the Belene site, an analysis was performed using revised seismic input data as well as two alternative foundation concepts (natural soil and soil exchange). The starting point for the analysis was the development of a suitable model of the coupled structures (base building, external building, containment, internal structure) and soil taking into account the real properties of the originally layered as well as the exchanged soil. The soil-structure effects were considered according to the analytical method employed, either through soil impedance (substructure method) or explicitly by a complex (direct method). On the basis of the results obtained by the two methods (substructure and direct method) the seismic safety of the complex structures for different foundation concepts was evaluated. By comparing the calculated structural response with the design spectra originally used for the design of components and systems the available safety margin was estimated

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.

Seismic margin review of the Maine Yankee Atomic Power Station: Fragility analysis

International Nuclear Information System (INIS)

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

1987-03-01

This Fragility Analysis is the third of three volumes for the Seismic Margin Review of the Maine Yankee Atomic Power Station. Volume 1 is the Summary Report of the first trial seismic margin review. Volume 2, Systems Analysis, documents the results of the systems screening for the review. The three volumes are part of the Seismic Margins Program initiated in 1984 by the Nuclear Regulatory Commission (NRC) to quantify seismic margins at nuclear power plants. The overall objectives of the trial review are to assess the seismic margins of a particular pressurized water reactor, and to test the adequacy of this review approach, quantification techniques, and guidelines for performing the review. Results from the trial review will be used to revise the seismic margin methodology and guidelines so that the NRC and industry can readily apply them to assess the inherent quantitative seismic capacity of nuclear power plants

Seismic margin review of the Maine Yankee Atomic Power Station: Fragility analysis

Energy Technology Data Exchange (ETDEWEB)

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

1987-03-01

This Fragility Analysis is the third of three volumes for the Seismic Margin Review of the Maine Yankee Atomic Power Station. Volume 1 is the Summary Report of the first trial seismic margin review. Volume 2, Systems Analysis, documents the results of the systems screening for the review. The three volumes are part of the Seismic Margins Program initiated in 1984 by the Nuclear Regulatory Commission (NRC) to quantify seismic margins at nuclear power plants. The overall objectives of the trial review are to assess the seismic margins of a particular pressurized water reactor, and to test the adequacy of this review approach, quantification techniques, and guidelines for performing the review. Results from the trial review will be used to revise the seismic margin methodology and guidelines so that the NRC and industry can readily apply them to assess the inherent quantitative seismic capacity of nuclear power plants.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

On the structural integrity evaluation about aged components

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

About one third of the nuclear power plants in Japan have been operated more than 30 years and flaws due to age-related degradation mechanisms have been detected in some components such as piping systems or core shrouds these years. Moreover, several severe earthquakes such as the Tohoku District - off the Pacific Ocean Earthquake or the Niigata-ken Chuetsu-oki Earthquake have struck some nuclear power plants in Japan recent years. Therefore, the structural integrity evaluation about nuclear installations and components considering seismic loads and aging mechanisms has become more and more important. In this study, several evaluation methods were proposed to assess the crack growth rate under the seismic loading conditions, to assess the failure conditions or the realistic failure capacities of the aged piping systems considering seismic or general loading conditions. Furthermore, analysis codes were developed considering aging mechanisms to carry out the integrity evaluation, or the failure probability evaluation which is useful in the seismic PSA evaluation. All of these assessment methods and analysis codes are being used and will be used more and more in the cross-check analyses or the safety reviews about nuclear installations and components. (author)

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

International Nuclear Information System (INIS)

Wingo, H.E.

1992-01-01

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

Seismic hazard assessment of Iran

Directory of Open Access Journals (Sweden)

M. Ghafory-Ashtiany

1999-06-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

The evaluation of a small capacity shell and tube ammonia evaporator

Energy Technology Data Exchange (ETDEWEB)

Garcia-Valladares, O.; Hernandez, J.I.; Best y Brown, R. [Centro de Investigacion en Energia de la UNAM, Morelos (Mexico); Gonzalez, J.C. [Universidad Autonoma de Campeche (Mexico). Programa CADETRAA

2003-12-01

The use of ammonia as refrigerant is widespread in vapour compression and ammonia/water absorption systems. Ammonia is not actually used in low capacity applications mainly because of the lack of economical available equipment. For this reason, the objective of this study is the numerical and experimental evaluation of a small capacity ammonia shell and tube evaporator with enhanced heat transfer surfaces. An experimental system to evaluate small capacity heat exchangers was developed. A shell and tube evaporator with external low fin tubes was successfully tested. The experimental uncertainty for the evaporator capacity has been estimated within {+-}5.5%. The experimental results were used to validate a heat exchanger numerical tool that predicts reasonably well the cooling capacity and load outlet temperatures. The methodology presented in this work can be applied to evaluate other refrigerants in similar shell and tube evaporators and to optimize the design of an evaporator for a specific application. (author)

Seismic gaps and plate tectonics: seismic potential for major boundaries

Energy Technology Data Exchange (ETDEWEB)

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

1979-01-01

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

Government and voluntary sector differences in organizational capacity to do and use evaluation.

Science.gov (United States)

Cousins, J Bradley; Goh, Swee C; Elliott, Catherine; Aubry, Tim; Gilbert, Nathalie

2014-06-01

Research on evaluation capacity is limited although a recent survey article on integrating evaluation into the organizational culture (Cousins, Goh, Clark, & Lee, 2004) revealed that interest in the topic is increasing. While knowledge about building the capacity to do evaluation has developed considerably, less is understood about building the organizational capacity to use evaluation. This article reports on the results of a pan-Canadian survey of evaluators working in organizations (internal evaluators or organization members with evaluation responsibility) conducted in 2007. Reliability across all constructs was high. Responses from government evaluators (N=160) were compared to responses from evaluators who work in the voluntary sector (N=89). The former were found to self-identify more highly as 'evaluators' (specialists) whereas the latter tended to identify as 'managers' (non-specialists). As a result, government evaluators had significantly higher self-reported levels of evaluation knowledge (both theory and practice); and they spent more time performing evaluation functions. However, irrespective of role, voluntary sector respondents rated their organizations more favorably than did their government sector counterparts with respect to the antecedents or conditions supporting evaluation capacity, and the capacity to use evaluation. Results are discussed in terms of their implications for evaluation practice and ongoing research. Copyright © 2013 Elsevier Ltd. All rights reserved.

Probabilistic seismic vulnerability and risk assessment of stone masonry structures

Science.gov (United States)

Abo El Ezz, Ahmad

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

Capacity Evaluation for IEEE 802.16e Mobile WiMAX

Directory of Open Access Journals (Sweden)

Chakchai So-In

2010-01-01

Full Text Available We present a simple analytical method for capacity evaluation of IEEE 802.16e Mobile WiMAX networks. Various overheads that impact the capacity are explained and methods to reduce these overheads are also presented. The advantage of a simple model is that the effect of each decision and sensitivity to various parameters can be seen easily. We illustrate the model by estimating the capacity for three sample applications—Mobile TV, VoIP, and data. The analysis process helps explain various features of IEEE 802.16e Mobile WiMAX. It is shown that proper use of overhead reducing mechanisms and proper scheduling can make an order of magnitude difference in performance. This capacity evaluation method can also be used for validation of simulation models.

Seismic analysis for the ALMR

International Nuclear Information System (INIS)

Tajirian, F.F.

1992-01-01

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

Effects of Irregular Bridge Columns and Feasibility of Seismic Regularity

Science.gov (United States)

Thomas, Abey E.

2018-05-01

Bridges with unequal column height is one of the main irregularities in bridge design particularly while negotiating steep valleys, making the bridges vulnerable to seismic action. The desirable behaviour of bridge columns towards seismic loading is that, they should perform in a regular fashion, i.e. the capacity of each column should be utilized evenly. But, this type of behaviour is often missing when the column heights are unequal along the length of the bridge, allowing short columns to bear the maximum lateral load. In the present study, the effects of unequal column height on the global seismic performance of bridges are studied using pushover analysis. Codes such as CalTrans (Engineering service center, earthquake engineering branch, 2013) and EC-8 (EN 1998-2: design of structures for earthquake resistance. Part 2: bridges, European Committee for Standardization, Brussels, 2005) suggests seismic regularity criterion for achieving regular seismic performance level at all the bridge columns. The feasibility of adopting these seismic regularity criterions along with those mentioned in literatures will be assessed for bridges designed as per the Indian Standards in the present study.

Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

International Nuclear Information System (INIS)

Conrads, T.J.

1993-06-01

In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP is comprised of a number of consultants known for their knowledge of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation Guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects which might employ the TSEP guidelines

Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances

International Nuclear Information System (INIS)

Conrads, T.J.

1993-01-01

In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the U.S. Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP comprises a number of consultants known for their understanding of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects that might employ the TSEP guidelines

Seismic design practices for power systems

International Nuclear Information System (INIS)

Schiff, A.J.

1991-01-01

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

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

International Nuclear Information System (INIS)

Li Shuian

2002-01-01

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

Multicomponent seismic applications in coalbed methane development

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Experimental study on the seismic performance of new sandwich masonry walls

Science.gov (United States)

Xiao, Jianzhuang; Pu, Jie; Hu, Yongzhong

2013-03-01

Sandwich masonry walls are widely used as energy-saving panels since the interlayer between the outer leaves can act as an insulation layer. New types of sandwich walls are continually being introduced in research and applications, and due to their unique bond patterns, experimental studies have been performed to investigate their mechanical properties, especially with regard to their seismic performance. In this study, three new types of sandwich masonry wall have been designed, and cyclic lateral loading tests were carried out on five specimens. The results showed that the specimens failed mainly due to slippage along the bottom cracks or the development of diagonal cracks, and the failure patterns were considerably influenced by the aspect ratio. Analysis was undertaken on the seismic response of the new walls, which included ductility, stiffness degradation and energy dissipation capacity, and no obvious difference was observed between the seismic performance of the new walls and traditional walls. Comparisons were made between the experimental results and the calculated results of the shear capacity. It is concluded that the formulas in the two Chinese codes (GB 50011 and GB 50003) are suitable for the calculation of the shear capacity for the new types of walls, and the formula in GB 50011 tends to be more conservative.

Approach of seismic upgrading in Kashiwazaki-Kariwa Nuclear Power Plant

International Nuclear Information System (INIS)

Sato, Hitoshi

2009-01-01

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

Evaluation of Fourier integral. Spectral analysis of seismic events

International Nuclear Information System (INIS)

Chitaru, Cristian; Enescu, Dumitru

2003-01-01

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

Improvement of seismic observation systems in JOYO

International Nuclear Information System (INIS)

Sumino, Kozo; Suto, Masayoshi; Tanaka, Akihiro

2013-01-01

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

Earthquake Ground Motion Measures for Seismic Response Evaluation of Structures

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-15

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

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

International Nuclear Information System (INIS)

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

1994-02-01

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

Normative Values for a Functional Capacity Evaluation

NARCIS (Netherlands)

Soer, Remko; van der Schans, Cees P.; Geertzen, Jan H.; Groothoff, Johan W.; Brouwer, Sandra; Dijkstra, Pieter U.; Reneman, Michiel F.

2009-01-01

Objective: To establish normative values for a functional capacity evaluation (FCE) of healthy working subjects. Design: Descriptive. Setting: Rehabilitation center. Participants: Healthy working subjects (N=701; 448 men, 253 women) between 20 and 60 years of age, working in more than 180

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.

An integrated approach to evaluate food antioxidant capacity.

Science.gov (United States)

Sun, T; Tanumihardjo, S A

2007-11-01

Many methods are available for determining food antioxidant capacity, which is an important topic in food and nutrition research and marketing. However, the results and inferences from different methods may vary substantially because each complex chemical reaction generates unique values. To get a complete and dynamic picture of the ranking of food antioxidant capacity, relative antioxidant capacity index (RACI), a hypothetical concept, is created from the perspective of statistics by integrating the antioxidant capacity values generated from different in vitro methods. RACI is the mean value of standard scores transformed from the initial data generated with different methods for each food item. By comparing the antioxidant capacity of 20 commonly consumed vegetables in the U.S. market that were measured with 7 chemical methods, we demonstrated that the RACI correlated strongly with each method. The significant correlation of RACI with an independent data set further confirmed that RACI is a valid tool to assess food antioxidant capacity. The key advantage of this integrated approach is that RACI is in a numerical scale with no units and has consistent agreement with chemical methods. Although it is a relative index and may not represent a specific antioxidant property of different food items, RACI provides a reasonably accurate rank of antioxidant capacity among foods. Therefore, it can be used as an integrated approach to evaluate food antioxidant capacity.

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

Science.gov (United States)

Merriam, M. K.

2005-12-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Lessons learned from NRC systematic evaluation program seismic review

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Ito, Hiroshi; Kitahara, Yoshihiro; Hirata, Kazuta

1986-01-01

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

Seismic load resistance of reinforcing steels in the as delivered condition and after corrosion - relevant material characteristics for performance evaluation

Energy Technology Data Exchange (ETDEWEB)

Moersch, Ing. Joerg [Max Aicher Engineering GmbH, Freilassing (Germany)

2016-10-15

This type of accelerated corrosion test was used to study the high number of test samples in due time. The corrosion phenomena obtained in salt spray testing deviate significantly from corrosion phenomena (pitting factor) obtained in practical conditions. Salt spray testing represents practical conditions for the more uniform corrosion as a result of a severe carbonation of the concrete and/or for higher chloride contents at the surface of the rebar. At low corrosion current densities the effect of pit depth on residual mechanical performance might be underestimated. Reinforced concrete (r.c.) buildings in seismic areas shall be designed to guarantee enough ductile resources as for example a sufficient rotational capacity to allow for load re-distribution. The rotational capacity is directly dependent on the ductility of the reinforcing steel which is generally expressed as elongation at maximum load (A+g{sub t}) and the hardening ratio (R{sub m}/R{sub e}). A direct testing of the seismic load resistance of reinforcing steels is not part of the construction product standards. Therefore it was decided by European Commission to introduce this performance requirement in the mandate for the revision of EN 10080:2005. In parallel to the standardization process a research project was carried out to deliver the scientific background.

Annual Hanford seismic report - fiscal year 1996

International Nuclear Information System (INIS)

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

1996-12-01

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

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)

Test Series 4: seismic-fragility tests of naturally-aged Exide EMP-13 battery cells

International Nuclear Information System (INIS)

Bonzon, L.L.; Hente, D.B.; Kukreti, B.M.; Schendel, J.; Tulk, J.D.; Janis, W.J.; Black, D.A.; Paulsen, G.D.; Aucoin, B.D.

1985-03-01

This report, the fourth in a test series of an extensive seismic research program, covers the testing of a 27-year old lead-antimony Exide EMP-13 cells from the recently decommissioned Shippingport Atomic Power Station. The Exide cells were tested in two configurations using a triaxial shake table: single-cell tests, rigidly mounted; and multicell (five-cell) tests, mounted in a typical battery rack. A total of nine electrically active cells was used in the two different cell configurations. None of the nine cells failed during the actual seismic tests when a range of ZPAs up to 1.5 g was imposed. Subsequent discharge capacity tests of five of the cells showed, however, that none of the cells could deliver the accepted standard of 80% of their rated electrical capacity for 3 hours. In fact, none of the 5 cells could deliver more than a 33% capacity. Two of the seismically tested cells and one untested, low capacity cell were disassembled for examination and metallurgical analyses. The inspection showed the cells to be in poor condition. The negative plates in the vicinity of the bus connections were extremely weak, the positive buses were corroded and brittle, negative and positive active material utilization was extremely uneven, and corrosion products littered the cells

Nuclear power plants seismic review programme in Spain

International Nuclear Information System (INIS)

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

1995-01-01

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

Energy-Based Design Criterion of Dissipative Bracing Systems for the Seismic Retrofit of Frame Structures

Directory of Open Access Journals (Sweden)

Gloria Terenzi

2018-02-01

Full Text Available Direct sizing criteria represent useful tools in the design of dissipative bracing systems for the advanced seismic protection of existing frame structures, especially when incorporated dampers feature a markedly non-linear behaviour. An energy-based procedure is proposed herein to this aim, focusing attention on systems including fluid viscous devices. The procedure starts by assuming prefixed reduction factors of the most critical response parameters in current conditions, which are evaluated by means of a conventional elastic finite element analysis. Simple formulas relating the reduction factors to the equivalent viscous damping ratio of the dampers, ξeq, are proposed. These formulas allow calculating the ξeq values that guarantee the achievement of the target factors. Finally, the energy dissipation capacity of the devices is deduced from ξeq, finalizing their sizing process. A detailed description of the procedure is presented in the article, by distinguishing the cases where the prevailing structural deficiencies are represented by poor strength of the constituting members, from the cases having excessive horizontal displacements. A demonstrative application to the retrofit design of a reinforced concrete gym building is then offered to explicate the steps of the sizing criterion in practice, as well as to evaluate the enhancement of the seismic response capacities generated by the installation of the dissipative system.

Dynamic behavior structural response and capacity evaluation of the standardized WWER-1000 nuclear power plants subjected to severe loading conditions

International Nuclear Information System (INIS)

Ambriashvili, Y.K.; Krutzik, N.J.

1993-01-01

In order to verify the structural capacity of standardized WWER-1000 MW nuclear power plants, comprehensive static and dynamic analyses were performed in cooperation between Siemens and Atomenergoprojekt. The main goal of these investigations was to perform of a number of seismic analyses of standardized WWER-1000 reactor buildings on the basis of 13 given seismological inputs, taking into account the local soil conditions at 17 different sites defined by in-situ investigations. The analyses were based on appropriate mathematical models (equivalent beam models as well as detailed spatial surface element models) of the coupled vibrating structures (base structure, outer structure, containment, inner structure) and of the layered soil. The analyses were mainly performed using the indirect method (substructure method). Based on the results of the seismic analysis as well as the results of static analysis (pressure and temperature due to LOCA, dead weight, prestressing) an assessment was made of the seismic safety of the containment and the reactor building. Using a complex 3-dimensional model of the structure and the soil, the influence of the flexibility of the basement structure on the structural response was also studied. The structural analyses of the WWER-1000 reactor building led to the conclusion that its design accounts well for the main factors governing the dynamic behavior of the building. The assessment of the forces acting in the structures shows that the bearing capacity of the analyzed building structure corresponds to an earthquake intensity of about 0.2 g to 0.25 g

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Seismic design of nuclear power plants - an assessment

International Nuclear Information System (INIS)

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

1976-01-01

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

Normative values for a functional capacity evaluation.

Science.gov (United States)

Soer, Remko; van der Schans, Cees P; Geertzen, Jan H; Groothoff, Johan W; Brouwer, Sandra; Dijkstra, Pieter U; Reneman, Michiel F

2009-10-01

Soer R, van der Schans CP, Geertzen JH, Groothoff JW, Brouwer S, Dijkstra PU, Reneman MF. Normative values for a functional capacity evaluation. To establish normative values for a functional capacity evaluation (FCE) of healthy working subjects. Descriptive. Rehabilitation center. Healthy working subjects (N=701; 448 men, 253 women) between 20 and 60 years of age, working in more than 180 occupations. Subjects performed a 2-hour FCE consisting of 12 work-related tests. Subjects were classified into categories based on physical demands according to the Dictionary of Occupational Titles. Means, ranges, SDs, and percentiles were provided for normative values of FCE, and a regression analysis for outcome of the 12 tests was performed. Normative FCE values were established for 4 physical demand categories. The normative values enable comparison of patients' performances to these values. If a patient's performance exceeds the lowest scores in his/her corresponding demand category, then the patient's capacity is very likely to be sufficient to meet the workload. Further, clinicians can make more precise return-to-work recommendations and set goals for rehabilitation programs. A comparison of the normative values can be useful to the fields of rehabilitation, occupational, and insurance medicine. Further research is needed to test the validity of the normative values with respect to workplace assessments and return-to-work recommendations.

Seismic and Power Generation Performance of U-Shaped Steel Connected PV-Shear Wall under Lateral Cyclic Loading

Directory of Open Access Journals (Sweden)

Hongmei Zhang

2014-01-01

Full Text Available BIPV is now widely used in office and residential buildings, but its seismic performance still remained vague especially when the photovoltaic (PV modules are installed on high-rise building facades. A new form of reinforced concrete shear wall integrated with photovoltaic module is proposed in this paper, aiming to apply PV module to the facades of high-rise buildings. In this new form, the PV module is integrated with the reinforced concrete wall by U-shaped steel connectors through embedded steel plates. The lateral cyclic loading test is executed to investigate the seismic behavior and the electric and thermal performance with different drift angles. The seismic behavior, including failure pattern, lateral force-top displacement relationship, and deformation capacity, was investigated. The power generation and temperature variation on the back of the PV module and both sides of the shear wall were also tested. Two main results are demonstrated through the experiment: (1 the U-shaped steel connectors provide enough deformation capacity for the compatibility of the PV module to the shear wall during the whole cyclic test; (2 the electricity generation capacity is effective and stable during this seismic simulation test.

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)

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

International Nuclear Information System (INIS)

Xu, J.; Graves, H.

1991-01-01

In the process of review and evaluation of licensing issues related to nuclear power plants, it is essential to understand the behavior of seismic loading, foundation and structural properties and their impact on the overall structural response. In most cases, such knowledge could be obtained by using simplified engineering models which, when properly implemented, can capture the essential parameters describing the physics of the problem. Such models do not require execution on large computer systems and could be implemented through a personal computer (PC) based capability. Recognizing the need for a PC software package that can perform structural response computations required for typical licensing reviews, the US Nuclear Regulatory Commission sponsored the development of a PC operated computer software package CARES (Computer Analysis for Rapid Evaluation of Structures) system. This development was undertaken by Brookhaven National Laboratory (BNL) during FY's 1988 and 1989. A wide range of computer programs and modeling approaches are often used to justify the safety of nuclear power plants. It is often difficult to assess the validity and accuracy of the results submitted by various utilities without developing comparable computer solutions. Taken this into consideration, CARES is designed as an integrated computational system which can perform rapid evaluations of structural behavior and examine capability of nuclear power plant facilities, thus CARES may be used by the NRC to determine the validity and accuracy of analysis methodologies employed for structural safety evaluations of nuclear power plants. CARES has been designed to operate on a PC, have user friendly input/output interface, and have quick turnaround. This paper describes the various features which have been implemented into the seismic module of CARES version 1.0

Seismic Performance of Composite Shear Walls Constructed Using Recycled Aggregate Concrete and Different Expandable Polystyrene Configurations

Directory of Open Access Journals (Sweden)

Wenchao Liu

2016-03-01

Full Text Available The seismic performance of recycled aggregate concrete (RAC composite shear walls with different expandable polystyrene (EPS configurations was investigated. Six concrete shear walls were designed and tested under cyclic loading to evaluate the effect of fine RAC in designing earthquake-resistant structures. Three of the six specimens were used to construct mid-rise walls with a shear-span ratio of 1.5, and the other three specimens were used to construct low-rise walls with a shear-span ratio of 0.8. The mid-rise and low-rise shear walls consisted of an ordinary recycled concrete shear wall, a composite wall with fine aggregate concrete (FAC protective layer (EPS modules as the external insulation layer, and a composite wall with sandwiched EPS modules as the insulation layer. Several parameters obtained from the experimental results were compared and analyzed, including the load-bearing capacity, stiffness, ductility, energy dissipation, and failure characteristics of the specimens. The calculation formula of load-bearing capacity was obtained by considering the effect of FAC on composite shear walls as the protective layer. The damage process of the specimen was simulated using the ABAQUS Software, and the results agreed quite well with those obtained from the experiments. The results show that the seismic resistance behavior of the EPS module composite for shear walls performed better than ordinary recycled concrete for shear walls. Shear walls with sandwiched EPS modules had a better seismic performance than those with EPS modules lying outside. Although the FAC protective layer slightly improved the seismic performance of the structure, it undoubtedly slowed down the speed of crack formation and the stiffness degradation of the walls.

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.

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

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Hassan Amirzehni

2016-12-01

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

Siting of nuclear desalination plants in Saudi Arabia: A seismic study

International Nuclear Information System (INIS)

Aljohani, M.S.; Abdul-Fattah, A.F.; Almarshad, A.I.

2005-01-01

This paper presents the selection criteria generally and seismic criteria specifically to select a suitable site in Saudi Arabia for a nuclear desalination plant. These criteria include geological, meteorological, cooling water supply discharge, transport infrastructure, population, electric grid, water network capacity, environmental impact and airport movement. The seismicity of the Arabian peninsula for the locations of seismic activity along the Red Sea and the Arabian Gulf coastlines from 1973 to 2000 was studied carefully. This study included towns and locations along the east and west coastlines and their distances from the seismic event site. The results showed that Rabigh City along the west coast of Saudi Arabia is a good site to build a nuclear desalination plant. This is because of the following reasons: good seismic stability; good weather statistics; no flooding; mild wave conditions; good supply and discharge; good transportation infrastructure; low population area; very close to the huge electric grid. (author)

Technical guidelines for the seismic safety re-evaluation at Eastern European NPPs

International Nuclear Information System (INIS)

Godoy, A.R.; Guerpinar, A.

2001-01-01

The paper describes one of the outcomes of the Engineering Safety Review Services (ESRS) that the IAEA provides as an element of the Agency's national, regional and interregional technical assistance and co-operation programmes and other extrabudgetary programmes to assess the safety of nuclear facilities. This refers to the establishment of detailed guidelines for conducting the seismic safety re-evaluation of existing nuclear power plants in Eastern European countries in line with updated criteria and current international practice. (author)

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.)

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)

Evaluation of Bearing Capacity of Strip Footing Using Random Layers Concept

Directory of Open Access Journals (Sweden)

Kawa Marek

2015-09-01

Full Text Available The paper deals with evaluation of bearing capacity of strip foundation on random purely cohesive soil. The approach proposed combines random field theory in the form of random layers with classical limit analysis and Monte Carlo simulation. For given realization of random the bearing capacity of strip footing is evaluated by employing the kinematic approach of yield design theory. The results in the form of histograms for both bearing capacity of footing as well as optimal depth of failure mechanism are obtained for different thickness of random layers. For zero and infinite thickness of random layer the values of depth of failure mechanism as well as bearing capacity assessment are derived in a closed form. Finally based on a sequence of Monte Carlo simulations the bearing capacity of strip footing corresponding to a certain probability of failure is estimated. While the mean value of the foundation bearing capacity increases with the thickness of the random layers, the ultimate load corresponding to a certain probability of failure appears to be a decreasing function of random layers thickness.

Seismic Risk of the Base Isolation System Protected by the Hard Stop

International Nuclear Information System (INIS)

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

2015-01-01

The concept of base isolation is to permit the deformation of isolator for absorbing seismic input wave from the ground. In a nuclear power plant design, allowable shear deformation of isolators should be enough to absorb the displacement response by extended design basis (EDB) ground motions. However isolators cannot resist over its displacement capacity. So, the clearance of hard stop (CHS) needs to be set between the response of base isolation system excited by the EDB ground motion and the displacement capacity of isolators. The isolation system must survive with high confidence in any seismic accident because it is a non-redundant system. Therefore, the CHS should be determined carefully based on the failure risk of base isolation system considering the uncertainties of earthquake responses and isolator capacities. In this research, the fragility curve of isolation system and its failure risk were estimated. The procedure to calculate the acceleration based fragility curve of the isolation system was developed. The fragility curve and failure risk for example case was estimated and its result was compared with different isolator capacities

Seismic Risk of the Base Isolation System Protected by the Hard Stop

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Han; Choi, In-Kil; Kim, Min Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The concept of base isolation is to permit the deformation of isolator for absorbing seismic input wave from the ground. In a nuclear power plant design, allowable shear deformation of isolators should be enough to absorb the displacement response by extended design basis (EDB) ground motions. However isolators cannot resist over its displacement capacity. So, the clearance of hard stop (CHS) needs to be set between the response of base isolation system excited by the EDB ground motion and the displacement capacity of isolators. The isolation system must survive with high confidence in any seismic accident because it is a non-redundant system. Therefore, the CHS should be determined carefully based on the failure risk of base isolation system considering the uncertainties of earthquake responses and isolator capacities. In this research, the fragility curve of isolation system and its failure risk were estimated. The procedure to calculate the acceleration based fragility curve of the isolation system was developed. The fragility curve and failure risk for example case was estimated and its result was compared with different isolator capacities.

Seismic hazard evaluation for major cities in Madagascar

International Nuclear Information System (INIS)

Razafindrakoto, Hoby N.T.; Rambolamanana, Gerard; Panza, Giuliano F.

2009-09-01