Seismic Qualification of Nuclear Plant Components Using PSHA-Based Spectra

2006 ◽  
Author(s):  
Divakar Bhargava ◽  
Keshab K. Dwivedy
Keyword(s):  
High Frequency ◽  
Nuclear Power ◽  
Energy Content ◽  
Hazard Analysis ◽  
Nuclear Plant ◽  
Mechanical Equipment ◽  
Piping System ◽  
Eastern United States ◽  
Damage Potential ◽  
Nuclear Plants

With the prospect of a revival of nuclear power industry after a long hiatus, there is an emphasis on designing the next breed of nuclear plants in the US using seismic spectra derived from a probabilistic seismic hazard analysis (PSHA). The methods available in guidance documents to establish Safe Shutdown Earthquake (SSE) spectral shapes at a site using PSHA invariably show that the risk-based spectra have high peaks, high zero period accelerations (ZPA), and significant energy content at higher frequencies when compared to the previous deterministic spectra at the same site. It is well known that earthquakes in Central and Eastern United States (CEUS) will typically contain some high frequency energy. While the early site permit applications and reactor supplier’s design certifications for new plants are expected to use the PSHA-based spectra for their seismic design, existing nuclear plants designed to deterministic spectra may also need to be reviewed for the probabilistic seismic spectra at their sites. This paper considers the implications of a probabilistic hazard spectrum for the seismic qualification of equipment and components for an operating plant and suggests a procedure for conducting a review. Amplification of ground spectra through nuclear plant structures and other intervening systems such as a piping system or an electrical cabinet are calculated using conventional linear dynamic analysis methods in much the same way as was done in the past for high frequency hydrodynamic loads in the Boiling Water Reactor (BWR) containments. Electrical and mechanical equipment, including devices such as relays that may be sensitive to high frequency vibratory loads are evaluated. While the spectral peaks at equipment mounting location are high at higher frequencies, the damage potential is considerably low. For an existing plant, a limited review of the previous seismic analyses and testing with the redefined seismic spectra concludes that the previous design has sufficient seismic margin. Implications of the PSHA based spectra for seismic qualification of equipment for new plants is not expected to be as severe as once believed. Additional assurance of safety can be obtained by updating or conducting a plant-specific seismic probabilistic risk analysis.

Effects of High Frequency Hydrodynamic Loads on Structural Integrity and Mechanical Operability of Valves

2015 ◽  
Author(s):  
Yigit Isbiliroglu ◽  
Cagri Ozgur ◽  
Evren Ulku ◽  
Nish Vaidya ◽  
Kristofor Paserba
Keyword(s):  
High Frequency ◽  
Structural Integrity ◽  
Energy Content ◽  
Hybrid Approach ◽  
Piping System ◽  
Seismic Loads ◽  
Damage Potential ◽  
Element Analysis ◽  
Piping Systems ◽  
Number Of Cycles

In-line valves are qualified for static as well as dynamic loads from seismic and hydrodynamic (HD) events. Seismic loads are generally characterized by frequency content less than about 33 Hz whereas HD loads may exhibit a broad range of frequencies greater than 33 Hz. HD loads may also result in spectral accelerations significantly in excess of those due to the design basis seismic events. Current regulatory guidelines do not specifically address the evaluation of equipment response to high frequency loading. This paper investigates the response of skid and line mounted valves of piping systems under HD loads by using several independent rigorous finite element analysis solutions for various piping system segments. It presents a hybrid approach for the evaluation of the response of valves to HD and seismic loads. The proposed approach significantly reduces the amount of individual analysis and testing needed to qualify the valves. First, valve responses are evaluated on the basis of displacements since HD loads are generally characterized by high frequencies and small durations. Second, the damage potential of the loads on the valve actuators is represented by the energy imparted to the actuator quantified in terms of Arias intensity. The rationale for using the energy content is based on the fact that damage due to dynamic loading is related not only to the amplitude of the acceleration response but also to the duration and the number of cycles over which this acceleration is imposed.

Conceptual Design of a 4 × 300 MW Modular Nuclear Plant

2011 ◽  
Author(s):  
Asko Vuorinen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Plant ◽  
Medium Size ◽  
Modular Construction ◽  
Construction Costs ◽  
University Of Technology ◽  
Nuclear Plants ◽  
Original Estimate ◽  
Under Construction

The Finnish companies have built four medium size nuclear power plants. In addition they have constructed two nuclear icebreakers and several floating power plants. The latest 1650 MWe nuclear power plant under construction Olkiluoto-3 has had many problems, which have raised the costs of the plant to €3500/kWe from its original estimate of €2000/kWe and constriction schedule from four to eight years. It is possible to keep the costs down and schedule short by making the plant in shipyard and transport it to site by sea. The plant could be then lifted to its place by pumping seawater into the channel. This kind of concept was developed by the author in 1991, when he was making his thesis of modular gas fired power plants in Helsinki University of Technology. The modular construction of nuclear plants has made in a form of two nuclear icebreakers, which Wa¨rtsila¨ Marine has built in Helsinki Shipyard. The latest modular nuclear plant was launched in 2010 in St Petersburg shipyard. One of the benefits of modular construction is a possibility to locate the plant under rock by making the transportation channels in tunnels. This will give the plant external protection for aircraft crash and make the outer containment unnecessary. The water channels could also be used as pressure suppression pools in case of venting steam from the containment. This could reduce the radioactive releases in case of possible reactor accidents. The two 440 MW VVER plants build in Finland had construction costs of €1600 /kWe at 2011 money. The author believes that a 1200 MW nuclear plant with four 300 MW units can be constructed in five years and with €3300/kW costs, where the first plant could be generating power within 40 months and next units with 6 month intervals.

Acoustic and Ultrasonic Signals as Diagnostic Tools for Check Valves

1993 ◽  
Vol 115 (2) ◽  
pp. 135-141 ◽  
Author(s):  
M. K. Au-Yang
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Operating Experience ◽  
Check Valve ◽  
Nuclear Regulatory Commission ◽  
Nuclear Plant ◽  
Diagnostic Tools ◽  
Piping System ◽  
Proper Training ◽  
Regulatory Commission

A typical nuclear plant has between 60 and 115 safety-related check valves ranging from 2 to 30 in. The majority of these valves control water flow. Recent studies done by the Institute of Nuclear Power Operations (INPO), Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC) found that many of these safety-related valves were not functioning properly. Typical problems found in these valves included disk flutter, backstop tapping, flow leakage, disk pin and hinge pin wear, or even missing disks. These findings led to INPO’s Significant Operating Experience Report (SOER, 1986), and finally, NRC generic letter 89-04, which requires that all safety-related check valves in a nuclear plant be regularly monitored. In response to this need, the industry has developed various diagnostic equipment to monitor and test check valves, using technologies ranging from acoustics and ultrasonics to magnetic—even radiography has been considered. Of these, systems that depend on a combination of acoustic and ultrasonic techniques (Au-Yang et al., 1991) are among the most promising for two reasons: these two technologies supplement each other, making diagnosis of the check valves much more certain than any single technology, and this approach can be made nonintrusive. The nonintrusive feature allows the check valves to be monitored and diagnosed without being disassembled or removed from the piping system. This paper shows that by carefully studying the acoustic and ultrasonic signatures acquired from a check valve, either individually or in combination, an individual with the proper training and experience in acoustic and ultrasonic signature analyses can deduce the structural integrity of the check valve with good confidence. Most of the conclusions are derived from controlled experiments in the laboratory where the diagnosis can be verified. Other conclusions were based on test data obtained in the field.

Fortum Participation in IAEA Benchmark for KKNPP7 RHR Piping Response NCO 2007 Event

2010 ◽  
Author(s):  
Pentti Varpasuo ◽  
Jukka Ka¨hko¨nen
Keyword(s):  
Numerical Simulation ◽  
Nuclear Power ◽  
Heat Removal ◽  
Tensile Failure ◽  
System Response ◽  
Engineering Practice ◽  
Mechanical Equipment ◽  
Piping System ◽  
Reactor Building ◽  
Residual Heat

In this report Fortum participation in a benchmark related to the residual heat removal (RHR) piping system response of the Kashiwazaki-Kariwa Nuclear Power Plant unit 7 (KKNPP7) of Tokyo Electric Power Company (TEPCO) in Japan, during the 16 July 2007 Niigataken-Chuetsu-Oki earthquake (NCO) is described. The goal of this benchmark is to conduct a comparison between different analytical techniques, as used in the usual engineering practice. Equipment behavior during the NCO earthquake constitutes an extensive database. There are no direct acceleration or displacement measurements on equipment. For this reason quantitative simulation comparison is not possible. Qualitative observations of the fuel tank sloshing and tank buckling are available and these can be used in numerical simulation benchmark. The benchmark is divided in three phases and it will be carried out during the years of 2009–2011. The number of participants in the benchmark is 25 organizations and institutions. Detailed goals of the benchmark as a whole can be summarized as follows: 1) Understanding of soil, structures and mechanical equipment response during the Niigataken-Chuetsu-Oki earthquake. 2) Simulation of equipment response for residual heat removal equipment. 3) Simulation of liquid sloshing in the fuel pools in reactor building. 4) Simulation of buckling phenomena of vertical, cylindrical yard tanks. In order to carry out this simulation task a joint model of the reactor building and the residual heat removal-system is developed. The main results of the numerical simulation will be the maximum values of the stresses in the most critical locations of the residual heat removal piping system. These values will be compared to known material properties of the piping such as yield and the tensile failure strain.

Vulnerability of Nuclear Power Plants to Voltage Sag Induced by Disturbances in the Transmission Grid: The Concept of Zone of Vulnerability for Voltage

2009 ◽  
Author(s):  
Nick Abi-Samra ◽  
Alberto Del Rosso ◽  
Frank Rahn
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Plant ◽  
Voltage Sag ◽  
Voltage Sags ◽  
Diesel Generator ◽  
Power Station ◽  
Transmission Grid ◽  
Nuclear Plants ◽  
Different Types

Nuclear plants are particularly sensitive to events on the grid that may lead to undervoltage on the auxiliary or safety buses. Momentary voltage dips can cause separation from offsite power and operation of emergency diesel generator. Grid disturbances that occur in certain areas or “zones” in the surrounding network, may affect the operation of the nuclear plant, while disturbances from outside these zones may be cause no threat to the nuclear plant. A zone inside which a nuclear power station would be vulnerable to events on the surrounding grid can be defined as the Zone of Vulnerability (ZoV). Different types of ZoVs can be defined depending on the nature of the vulnerability being considered. This paper deals specifically with the Zone of Vulnerability associated with the risk of voltage sags in the safety and auxiliary for safety and auxiliary buses of nuclear plants induced by faults in the power grid. The paper first introduces the concept of ZoV_v and its importance for nuclear plants security. A methodology for determining the ZoV_v is then described.

The Role of Displacement-Controlled Stresses in Critical Flaw Size Determination for Piping Systems

2008 ◽  
Author(s):  
Peter C. Riccardella ◽  
Paul Hirschberg ◽  
Ted Anderson ◽  
Greg Thorwald ◽  
Eric Scheibler
Keyword(s):  
Stainless Steel ◽  
Thermal Expansion ◽  
Nuclear Power ◽  
Power Plants ◽  
Bending Moment ◽  
Flaw Size ◽  
Piping System ◽  
Nuclear Plants ◽  
Piping Systems ◽  
Critical Flaw

A debate has long ensued in ASME Subcommittee XI regarding the need to include displacement-controlled (secondary) stresses in critical flaw size calculations for austenitic weldments. There is general agreement that inclusion of secondary stresses is not necessary for highly ductile piping materials such as wrought stainless steel and high nickel alloys. However, some stainless steel weldments are classified as “low-toughness” because, although not considered brittle, they exhibit lower toughness than wrought stainless steel. The Code requires the inclusion of global secondary stresses, such as piping thermal expansion loads, in critical flaw size calculations for such weldments, albeit at reduced safety factors. The Code requirements are less clear for dissimilar metal weldments, such as Alloy 82/182, which were often used for ferritic nozzle to safe-end welds in nuclear power plants, and which have proven in service to be susceptible to a form of stress corrosion cracking. Analyses are presented in this paper that shed additional light on the subject. Finite element analyses (FEA) of a straight pipe with a through-thickness crack were used to determine the effect on bending moment and crack driving force due to an imposed end rotation. Moment and J-integral knock-down factors are computed for a range of crack sizes for two different pipe lengths. Piping analyses are also presented for two typical PWR surge lines, which are among the highest secondary stress locations in U.S. nuclear plants. These analyses predict the maximum rotation at the surge nozzle that could be produced by the secondary loads (anchor movement + thermal expansion + stratification), and compare that to rotations that were sustained in full scale pipe tests containing large complex cracks. The analyses demonstrate that secondary loads would be substantially reduced prior to fracture of a cracked weldment, and that they are therefore of reduced significance in critical flaw size calculations. A general method for estimating the effect of secondary loads on pipe fracture as a function of relative piping system and crack section stiffness is suggested.

A New Paradigm: Cycling Operations at Nuclear Power Plants in the United States

2014 ◽  
Author(s):  
Douglas Hilleman ◽  
Nikhil Kumar ◽  
Steven Lefton
Keyword(s):  
United States ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The United States ◽  
Nuclear Plant ◽  
New Paradigm ◽  
Load Following ◽  
Nuclear Plants ◽  
Relative Damage

Nuclear power plants are no longer immune to cycling operation. While certain nuclear power plants in Europe have been performing load following operation, this type of operation has largely been avoided in the United States. Due to increasing contribution of nuclear generation in the mix, European operators were forced to make modifications to increase the maneuverability of their nuclear generation assets. However, in the United States, nuclear generation is still a relatively smaller contributor (19%). Still, with rapid increase in renewable generation, some nuclear plants are being asked to operate at reduced power and cycle to lower power levels. With most future renewable integration studies advocating for increased flexibility on the grid, nuclear generation maneuverability will allow system operators with another resource to mitigate system costs. This paper presents the results of a detailed study of a 1,150 MW boiling water reactor nuclear plant when cycled to low loads. The authors present the relative damage of cycling to various reduced power levels 80% to 15% power levels compared to a cold startup and shutdown of a nuclear plant. An assessment was made of the systems that had fatigue damage and costs. We also discuss some of the limitations of cycling that a nuclear plant has and present and discuss recommendations to reduce damage and costs.

scholarly journals A ground motion logic tree for seismic hazard analysis in the stable cratonic region of Europe: regionalisation, model selection and development of a scaled backbone approach

2020 ◽  
Vol 18 (14) ◽  
pp. 6119-6148
Author(s):  
Graeme Weatherill ◽  
Fabrice Cotton
Keyword(s):  
United States ◽  
Seismic Hazard ◽  
Ground Motion ◽  
Epistemic Uncertainty ◽  
Hazard Analysis ◽  
Seismic Hazard Analysis ◽  
Eastern United States ◽  
Logic Tree ◽  
Short Period ◽  
Northeastern Europe

Abstract Regions of low seismicity present a particular challenge for probabilistic seismic hazard analysis when identifying suitable ground motion models (GMMs) and quantifying their epistemic uncertainty. The 2020 European Seismic Hazard Model adopts a scaled backbone approach to characterise this uncertainty for shallow seismicity in Europe, incorporating region-to-region source and attenuation variability based on European strong motion data. This approach, however, may not be suited to stable cratonic region of northeastern Europe (encompassing Finland, Sweden and the Baltic countries), where exploration of various global geophysical datasets reveals that its crustal properties are distinctly different from the rest of Europe, and are instead more closely represented by those of the Central and Eastern United States. Building upon the suite of models developed by the recent NGA East project, we construct a new scaled backbone ground motion model and calibrate its corresponding epistemic uncertainties. The resulting logic tree is shown to provide comparable hazard outcomes to the epistemic uncertainty modelling strategy adopted for the Eastern United States, despite the different approaches taken. Comparison with previous GMM selections for northeastern Europe, however, highlights key differences in short period accelerations resulting from new assumptions regarding the characteristics of the reference rock and its influence on site amplification.

Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC

2008 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Nuclear Power ◽  
Design Evaluation ◽  
Piping System ◽  
Design Verification ◽  
Intensive Study ◽  
Non Linear ◽  
Piping Systems ◽  
Design Requirements

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. This paper attempts first to categorize the design verification according to ASME III into the linear design and non-linear design verifications. Thereafter, the corresponding design requirements, in particular, those non-linear design requirements, are reviewed and examined in detail. The emphasis is placed on our view on several formulations and design requirements in ASME III when applied to nuclear power piping systems that are currently under intensive study in Sweden.

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