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For improving the seismic performance of the nuclear power plant (NPP) piping system, attempts have been made to apply a dynamic absorber (DA). However, the current piping DA design method is limited because it cannot provide the globally optimum values for the target design seismic loading. Therefore, this study proposes a seismic time history analysis-based DA optimal design method for piping. To this end, the Kriging approach is introduced to reduce the numerical cost required for seismic time history analyses. The appropriate design of the experiment method is used to increase the efficiency in securing response data. A gradient-based method is used to efficiently deal with the multi-dimensional unconstrained optimization problem of the DA optimal design. As a result, the proposed method showed an excellent response reduction effect in several responses compared to other optimal design methods. The proposed method showed that the average response reduction rate was about 9% less at the maximum acceleration, about 5% less at the maximum value of the response spectrum, about 9% less at the maximum relative displacement, and about 4% less at the maximum combined stress compared to existing optimal design methods. Therefore, the proposed method enables an effective optimal DA design method for mitigating seismic response in NPP piping in the future.

As a part of round robin analysis to develop a finite element elastic-plastic seismic analysis procedure for nuclear safety class 1 components, a series of parametric analyses was carried out on the simulated pressurizer surge line system model to investigate sensitivity of the analysis results to finite element analysis variables. The analysis on the surge line system model considered dynamic effect due to the seismic load corresponding to PGA 0.6 g and elastic-plastic material behavior based on the Chaboche combined hardening model. From the parametric analysis results, it was found that strains such as accumulated equivalent plastic strain and equivalent plastic strain are more sensitive to the analysis variables than von Mises effect stress. The parametric analysis results also identified that finite element density and ovalization option in the elbow elements have more significant effect on the analysis results than the other variables.

The purpose of this study is to reduce seismic responses of an actual nuclear piping system using a tuned mass damper (TMD) device. A numerical piping model was developed and validated based on shaking table test results with actual nuclear piping. A TMD for nuclear piping was newly devised in this work. A TMD shape design suitable for nuclear piping systems was conducted, and its operating performance was verified after manufacturing. The response reduction performance of the developed TMD under earthquake loading on actual piping was investigated. Results confirmed that, on average, seismic response reduction rates of 34% in the maximum acceleration response, 41% in the root mean square acceleration response, and 57% in the spectral acceleration response were shown through the TMD application. This developed TMD operated successfully within the seismic response reduction rate of existing TMD optimum design values. Therefore, the developed TMD and dynamic interpretation help improve the nuclear piping's seismic performance.

The main purpose of this study is to perform shaking table tests to validate the inelastic seismic analysis method applicable to pressure-retaining metal components in nuclear power plants (NPPs). To do this, the test mockup was designed and fabricated to be able to describe the hot leg surge line nozzle with a piping system, which is known to be one of the seismically fragile components in nuclear steam supply systems (NSSS). The used input motions are the displacement time histories corresponding to the design floor response spectrum at an elevation of 136 ft in the in-structure building in NPPs. Two earthquake levels are used in this study. One is the design-basis safe shutdown earthquake level (SSE, PGA = 0.3 g) and the other is the beyond-design-basis earthquake level (BDBE, PGA = 0.6 g), which is linearly scaled from the SSE level. To measure the inelastic strain responses, five strain gauges were attached at the expected critical locations in the target nozzle, and three accelerometers were installed at the shaking table and piping system to measure the dynamic responses. From the results of the shaking table tests, it was found that the plastic strain response at the target nozzle and the acceleration response at the piping system were not amplified by as much as two times the input earthquake level because the plastic behavior in the piping system significantly contributed to energy dissipation during the seismic events. To simulate the test results, elastoplastic seismic analyses with the well-known Chaboche kinematic hardening model and the Voce isotropic hardening model for Type 316 stainless steel were carried out, and the results of the principal strain and the acceleration responses were compared with the test results. From the comparison, it was found that the inelastic seismic analysis method can give very reasonable results when the earthquake level is large enough to invoke plastic behavior in nuclear metal components.

The main purpose of this study is to investigate the feasibility of the seismic fragility analysis (FA) with the strain-based failure modes for the nuclear metal components retaining pressure boundary. Through this study, it is expected that we can find analytical ways to enhance the high confidence of low probability of failure (HCLPF) capacity potentially contained in the conservative seismic design criteria required for the nuclear metal components. Another goal is to investigate the feasibility of the seismic FA to be used as an alternative seismic design rule for beyond-design-basis earthquakes. To do this, the general procedures of the seismic FA using the inelastic seismic analysis for the nuclear metal components are investigated. Their procedures are described in detail by the exampled calculations for the surge line nozzles connecting hot leg piping and the pressurizer, known as one of the seismic fragile components in NSSS (Nuclear Steam Supply System). To define the seismic failure modes for the seismic FA, the seismic strain-based design criteria, with two seismic acceptance criteria against the ductile fracture failure mode and fatigue-induced failure mode, are used in order to reduce the conservatism contained in the conventional stress-based seismic design criteria. In the exampled calculation of the inelastic seismic strain response beyond an elastic regime, precise inelastic seismic analyses with Chaboche’s kinematic and Voce isotropic hardening material models are used. From the results of the seismic FA by the probabilistic approach for the exampled target component, it is confirmed that the approach of the strain-based seismic FA can extract the maximum seismic capacity of the nuclear metal components with more accurate inelastic seismic analysis minimizing the number of variables for the components.

In this study, a small laminated rubber bearing (LRB) using near-natural rubber is designed as a representative model of equipment seismic isolator to improve the seismic performance of the nuclear power plant (NPP). Some equipment of high safety importance is chosen to apply the seismic isolation using LRBs, and its design variables in consideration are reviewed to reduce seismic response. In comparison with the large LRBs for building structures, the control of design parameters for small LRBs, such as natural frequencies and shape factor, is very limited by the difficulties of fabrication. The paper discusses how they are updated and changed through the performance testing process.

This paper presents a design specification of the small-sized lead inserted laminated rubber bearing (LRB) for application to nuclear component seismic isolation and describes the results of test verification on design performance parameters such as effective horizontal stiffness, equivalent viscous damping ratio, design seismic isolation frequency, and ultimate shear deformation. To do this, two types of LRB, having the same vertical design load of 10 kN but with different shape factors, are designed, fabricated, and tested by the quasi-static procedures. To determine the effective horizontal stiffness and the equivalent damping value from the test results, the new method is proposed and compared with the methods of the ASCE and ISO standards in case that the tangential stiffness curve is not linear in tests. From the comparison between tests and design specifications in the performance parameters, it was found that the design specifications developed in this paper are in a good agreement with the test results. Furthermore, the target design shear deformation limits are confirmed to have sufficient design margins in ultimate shear deformation tests.

In this paper, the inelastic material models for Type 316H stainless steel, which is one of the principal candidate materials for elevated temperature design of the advanced high temperature reactors (HTRs) pressure retained components, are investigated and the required material parameters are identified to be used for both elasto-plastic models and unified viscoplastic models. In the constitutive equations of the inelastic material models, the kinematic hardening behavior is expressed with the Chaboche model with three backstresses, and the isotropic hardening behavior is expressed by the Voce model. The required number of material parameters is minimized to be ten in total. For the unified viscoplastic model, which can express both the time-independent plastic behavior and the time-dependent viscous behavior, the constitutive equations have the same kinematic and isotropic hardening parameters of the elasto-plastic material model with two additional viscous parameters. To identify the material parameters required for these constitutive equations, various uniaxial tests were carried out at isothermal conditions at room temperature and an elevated temperature range of 425&ndash, 650 &deg, C. The identified inelastic material parameters were validated through the comparison between tests and calculations.

In this paper, the feasibility study to develop strain-based seismic design criteria applicable for the components of nuclear power plants are carried out as an alternative rule to the current stress-based criteria. To do this, two acceptance criteria are investigated through the detailed example of an application for the surge line nozzles in a nuclear steam supply system, which are known as one of the seismic fragile components in nuclear power plants. These strain-based seismic design criteria are primarily to prevent two types of failure modes, such as a ductile fracture and a cyclic fatigue-induced damage due to continuous large amplitude cyclic loads during seismic event. Through the example problem, the required procedures are described step-by-step with calculations of an accumulated plastic strain, triaxiality factor by the elasto-plastic seismic analysis using the finite element method. For a precise inelastic seismic analysis, the Chaboche kinematic and Voce isotropic hardening material parameters are identified by the test data and used for an inelastic material model. The results by the strain-based criteria are compared with those by the ASME (American Society of Mechanical Engineers) stress-based design criteria for a service level D limits. From the study, it is expected that the strain-based seismic design method investigated in this paper will be beneficial for the nuclear components, especially when the design basis earthquakes are large enough to cause severe plastic strains at a critical location.

In a nuclear power plant, it is essential to improve the seismic safety of the piping system for the coolant transfer to cool the high temperature caused by the nuclear reaction. Under this background, this study makes two major contributions. The first is that though tuned mass dampers (TMDs) were originally used only to reduce the vibration of piping itself, through this research, it was first proved that it had a positive effect on the improvement of the seismic performance of nuclear piping systems. Additionally, this study proposed a design approach that effectively obtains the optimal design values of TMDs associated with seismic performance. In order to effectively derive the TMD optimum design values, we not only utilized the existing TMD optimum design formula, but also additionally proposed a frequency response analysis-based TMD optimal design method. As a result, it was seen that primary responses of system were significantly reduced under the input seismic load due to the use of TMDs for the piping system. It was also confirmed that the use of the existing TMD formula brought about a similar degree of response reduction effect, while it was possible to get the improved effect when using the proposed method.

In this paper, the seismic analysis modeling technologies for sodium-cooled fast reactor (SFR) are presented with detailed descriptions for each structure, system and component (SSC) model. The complicated reactor system of pool type SFR, which is composed of the reactor vessel, internal structures, intermediate heat exchangers, primary pumps, core assemblies, and core support structures, is mathematically described with simple stick models which can represent fundamental frequencies of SSC. To do this, detailed finite element analyses were carried out to identify fundamental beam frequencies with consideration of fluid added mass effects caused by primary sodium coolant contained in the reactor vessel. The calculation of fluid added masses is performed by detailed finite element analyses using FAMD computer program and the results are discussed in terms of the ways to be considered in a seismic modeling. Based on the results of seismic time history analyses for both seismic isolation and nonisolation design, the functional requirements for relative deflections are discussed, and the design floor response spectra are proposed that can be used for subsystem seismic design.

For elevated temperature design evaluations by the ASME-NH rules, the most important material data is the isochronous stress-strain curves, which can provide design creep information. The main purpose of this paper is to investigate appropriate formulations to be able to generate the isochronous stress-strain curves and implement it to the computer program which is coded the ASME-NH design evaluation procedures. To do this, formulations by the strain-time relationship are investigated in detail and the sensitivity studies for rapid initial transient creep contributions, slower and longer transient creep contribution, and secondary creep contributions are carried out for type 316 austenitic stainless steel. From the results of this study, it is found that the strain-time relationship formulations can well describe the isochronous stress-strain curves with the transient creep contributions.

The preliminary structural analyses for an advanced burner test reactor (ABTR) internal structure are performed with regard to the elevated temperature design and the seismic design. In the elevated temperature design, three load cases enveloped from ABTR duty cycle events are studied for the structural integrity evaluation according to the ASME-NH design rule. The resulting stresses and strains are compared with the allowable design and service limits. An elastic approach, a simplified inelastic approach and a creep fatigue analysis are performed to provide a quantitative assessment of the deformations and strains. In the seismic design, the response spectrum analyses including the fluid-structure interaction effect are performed for isolated and non-isolated conditions. From the evaluation results, it is found that the seismic isolation capable of reducing the seismic input transmitted to the foundations is needed to satisfy the allowable design limits. For the isolated condition, three design cases for the variation of the wall thickness as 1.5 cm, 2 cm and 3 cm are studied for the design modification.

We evaluated the buckling limit of a conceptually designed reactor vessel of the Advanced Burner Test Reactor (ABTR) for a horizontal safe shutdown earthquake (SSE) seismic load. In this evaluation, both seismic isolation and non-isolation designs were considered for thin reactor vessels subjected to elevated temperature services. For calculating the buckling load, two kinds of methods, a numerical simulation method using finite element analysis and an evaluation formula driven from a theoretical basis, were used. To consider the material aging effect caused by a 60-year design lifetime and 510°C normal operating temperature, an isochronous stress-strain curve corresponding to these conditions was used for the nonlinear elastic-plastic buckling analysis method. From the evaluation results of the buckling load, it was found that plasticity behavior significantly affects the buckling strength, but that the initial geometrical imperfections have little effect. Also, the non-seismic isolation design does not satisfy the buckling limit rules of the ASME BPV Section III, Subsection NH, but the seismic isolation design does satisfy it with sufficient margins.

In this paper, the material parameters of plastic and viscoplastic constitutive equations for modified 9Cr–1Mo steel are developed for various isothermal conditions to support inelastic analysis for a sodium-cooled fast reactor. To do this, the material parameters related with the elastoplastic behaviour are identified with uniaxial cyclic test data by performing computer simulations, which use the combined Chaboche model including the kinematic hardening rule and the isotropic softening rule. The viscous parameters are identified from uniaxial stress relaxation test data through computer simulations with the pre-determined elastoplastic material parameters. Sensitivity studies are performed for the material parameters to investigate cyclic inelastic behaviour and stress relaxation during a hold time. From the comparison between the tests and the simulations, it is expected that the identified material parameters of the plastic and viscoplastic constitutive equations can accurately express the material characteristics of modified 9Cr–1Mo steel sufficiently well to be used for inelastic analysis.

In this paper, creep-fatigue damage under elevated temperatures is investigated for a tube sheet-to-shell structure, which is one of the main structures under Gen-IV class 1 components. To do this, detailed step-by-step procedures, including the elastic structural analysis and the ASME-NH code application, are described for a defined representative load cycle. From the sensitivity studies for various design parameters, such as hold time duration, shell thickness, and operating temperature, it is found that a reduction of thickness can decrease the thermal bending stresses, but the negative effect is that it may increase the primary stress and enhance the creep damage. The normal operating temperature is the most significant parameter in the creep-fatigue design.

An assessment of creep-fatigue crack initiation and growth for a 316L stainless steel structure has been carried out according to the current (2007 edition) and previous (2002 edition) versions of the French RCC-MR A16 procedure. Some significant changes have been made in terms of the formulae and material properties, which may cause big differences in the assessment. In this study, the changes in the A16 guide have been quantified for a 316L austenitic stainless steel structure, and the assessment results were compared with those of the observed images from a structural test for a welded component.

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

This study deals with the generation of isochronous stress-strain curves for modified 9Cr-1Mo steel. Material property data (sy, D, and m) from tensile tests and strain-time curves from creep tests were obtained at 550 °C. On the basis of the experimental data, the creep curves were characterized by Garofalo’s creep model. The three parameters of P1, P2 and P3 in Garofalo’s model could be optimized properly by a nonlinear least square fitting analysis. The three parameters showed a good stress dependency with a linear relation. However, the P3 parameter, which represents the steady state creep rate, exhibited two-slope behavior with different stress exponents at a transient stress of about 250 MPa. The results verify that the long-term creep curves for G91 steel can be modeled by Garofalo’s model with only several short-term creep data. When the modeled creep curves are used, long-term isochronous curves of up to 105 h can be successfully constructed.

In this paper, a method to consider temperature dependent material properties when using the Green's function method is proposed by using a numerical weight function approach. This is verified by using detailed finite element analyses for a pressurizer spray nozzle with various assumed thermal transient load cases. From the results, it is found that the temperature dependent material properties can significantly affect the maximum peak stresses and the proposed method can resolve this problem with the weight function approach. Finally, it is concluded that the temperature dependency of the material properties affects the maximum stress ranges for a fatigue evaluation. Therefore, it is necessary to consider this effect to monitor fatigue damage when using a Green's function method for the real operating conditions in a nuclear power plant.

In this paper, we discuss an ASME-NH program that has been developed to overcome the complexity and costs arising from the real application of the ASME-NH rules by hand calculations for class 1 nuclear facility component design for elevated temperature operations. A computerized program is described for implementing all the assessment procedures such as the time-dependent primary stress limits, total accumulated creep-ratcheting strain limits, and the creep–fatigue damage limits by the elastic and inelastic analysis methods complying with the ASME-NH rules. As an example application, a preliminary structural integrity evaluation for a high-temperature reactor vessel design of a typical lead-cooled reactor is described.

In LMR (Liquid Metal Reactor) core seismic behavior, the fluid coupling between closely spaced hexagons is known to have a very strong effect on the natural frequencies and the seismic responses. In this paper, an experimental study is carried out to investigate the fluid coupling effects with reduced scale mock-ups with three types of hexagonal duct systems and three types of input seismic motions such as the ElCentro 1940 time history, the artificial time history (Reg. 1.60), and the Kobe time history by using a one-dimensional shaking table facility. The results of the experimental studies are compared and discussed with those of the analyses performed by using the SAC-CORE code, which implements a numerical algorithm of the CFAM (Consistent Fluid Added Mass) matrix proposed in a previous paper. From the results, it is found that the fluid coupling between closely spaced hexagons becomes stronger with an increasing number of the neighboring hexagonal ducts, and this causes the natural frequency to be significantly changed and it affects the seismic time history responses too.

In this paper, a simple seismic analysis model of the KALIMER-600 sodium-cooled fast reactor selected to be the candidate of the GEN-IV reactor is developed. By using this model, the seismic time history analysis is carried out to investigate the feasibilities of a seismic isolation design. The developed simple seismic analysis model includes the reactor building, reactor system,, IHTS piping system, steam generator, and seismic isolators. The dynamic characteristics of the simple seismic model are verified with the detailed 3-dimensional finite element analysis for each part of the KALIMER-600 system. By using the developed simple seismic model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design are performed for the artificial time history of a SSE (Safe Shutdown Earthquake) 0.3g. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

In this paper, the inelastic constitutive models for the simulations of the cyclic softening behavior of the modified 9Cr-1Mo steel, which has a significant cyclic softening characteristic especially in elevated temperature regions, are investigated in detail. To do this, the plastic modulus, which primarily governs the calculation scheme of the plasticity, is formulated for the inelastic constitutive models such as the Armstrong-Frederick model, Chaboche model, and Ohno-Wang model. By implementing the extracted plastic modulus and the consistency conditions into the computer program, the inelastic constitutive parameters are identified to present the best fit of the uniaxial cyclic test data by strain-controlled simulations. From the computer simulations by using the obtained constitutive parameters, it is found that the Armstrong-Frederick model is simple to use but it causes significant overestimated strain results when compared with the Chaboche and the Ohno-Wang models. And from the ratcheting simulation results, it is found that the cyclic softening behavior of the modified 9Cr-1Mo steel can invoke a ratcheting instability when the applied cyclic loads exceed a certain level of the ratchet loading condition.

In this paper, an inelastic constitutive equation of the Chaboche model, which can simulate ratcheting behaviour at high temperatures, is investigated for modified 9Cr–1Mo steel. The plastic modulus, which primarily governs the calculation scheme for plasticity, is derived for the Chaboche model and implemented into a computer program with the radial return algorithm to compute the cyclic backstress components. With the cyclic test data of modified 9Cr–1Mo steel, the Chaboche three-decomposed constitutive parameters of the kinematic hardening rule and the isotropic softening rule, which controls the cyclic softening characteristic of modified 9Cr–1Mo steel, are identified by a simple method. By using the identified constitutive parameters, the ratcheting behaviour of modified 9Cr–1Mo steel is simulated to investigate the effects of mean stress levels at a certain stress amplitude and ratcheting parameter's values. In particular, the progressive deformation instability phenomena due to the cyclic softening characteristics of modified 9Cr–1Mo steel are investigated.

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor – Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid–structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules.

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500°C and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

In this paper, a numerical application algorithm for applying the CFAM (Consistent Fluid Added Mass) matrix for a core seismic analysis is developed and applied to the 7-ducts core system to investigate the fluid effects on the dynamic characteristics and the seismic time history responses. To this end, three cases such as the in-air condition, the in-water condition without the fluid coupling terms, and the in-water condition with the fluid coupling terms are considered in this paper. From modal analysis, the core duct assemblies revealed strongly coupled out-of-phase vibration modes unlike the other cases with the fluid coupling terms considered. From the results of the seismic time history analysis, it was also verified that the fluid coupling terms in the CFAM matrix can significantly affect the impact responses and the seismic displacement responses of the ducts.