Abstract At ageing power plants, local thinning of pipework or vessel is unavoidable due to erosion/corrosion or other reasons such as flow accelerated corrosion (FAC) — one of the common degradation mechanisms in pipework of nuclear power plant. Local thinning reduces the structure strength, resulting in crack initiation from the corrosion pit or welding defect when subject to cyclic loading. General practice is to use the minimum thickness of the thinned area to calculate both limit load and stress intensity factor (SIF) in performing Engineering Critical Assessment (ECA) using Failure Assessment Diagram (FAD). Using the minimum thickness is normally overly conservative as it assumes that thinning occurs grossly instead of locally, leading to unnecessary early repair/replacement and cost. Performing cracked body finite element analysis (FEA) can provide accurate values of limit load and SIF, but it is time consuming and impractical for daily maintenance and emergent support. To minimise the conservatisms and provide a guidance for the assessment of locally thinned pipework or vessel using existing handbook solutions, a study was carried out by the authors on the effect of local thinning on limit loads. The study demonstrates that local thinning has significant effect on limit load if the thinning ratio of thinning depth to original thickness is larger than 25%. It concluded that the limit load solutions given in handbooks (such as R6 or the net section method) are overly conservative if using the minimum local thickness and non-conservative if using the nominal thickness. This paper discusses the effect of local thinning on SIFs of internal/external defects using cracked body finite element method (FEM). The results are compared with R6 weight function SIF solutions for a cylinder. A modified R6 SIF solution is proposed to count for the effect of local thinning profile. Along with the previous published paper on limit load it provides comprehensive understanding and guidance for fracture assessment of the local thinned pipework and vessel.

Abstract Fatigue is identified as a significant degradation mode that affects nuclear power plants world-wide. Several international codes and standards (ASME, RCC-M, JSME, etc...) offer rules to predict its damaging effect on the locations of the various components of an NPP. These rules, which ensure conservatism and safe operation, have grown in complexity over the years because they have integrated R&D results showing aggravating effects that were not included in the original analyses (such as Environmental Assisted Fatigue[1][2]) but also because an economically viable design of components has required optimization and refinement of mechanical assessment methods. CNNC/NPIC has been following carefully the recent evolutions in the fatigue rules and has today finalized an in-house software enabling the evaluation of fatigue per ASME and RCC-M rules, with integration of environmental effects. On the other hand, EDF has been developing since 1989 its own in-house FEA code baptizedCode_Aster that is included in the Salome-Meca mechanical package. Salome-Meca is open-access and can be used freely by international users. Within Code_Aster, the fatigue postprocessor offers a span of criteria (Dang Van, Stress Intensity, etc...) to pick and choose from and even offers the possibility to make up owns one fatigue criteria. It also offers the possibility to post-process fatigue according to the RCC-M rules (POST_RCCM operator). It has been recently updated to perform industrial calculations integrating environmental fatigue[2][3]. Both entities have come to agree that validating a fatigue computer code is not an easy task. On the one hand, the full validation using hand calculations would be a highly tedious effort, given the technicality and the multiple choices to make along the various steps of the fatigue analysis. On the other hand, there are no experiments today which enable to directly lead a benchmark calculation to validate fatigue numerical results. In consequence, an accepted way of validating a code is to perform a benchmark analysis against another industrial fatigue code. CNNC/NPIC and EDF R&D China have therefore launched an effort to benchmark their respective codes, with the final objective of progressing together towards safe structural assessment practical methods for their power plants components.

Abstract Fluidelastic instability (FEI) is well known to be a critical flow-induced vibration concern for the integrity of the tubes in nuclear steam generators. Traditionally, this has been assumed to occur in the direction transverse to the direction of flow but the tube failures at San Onofre Nuclear Generating Station (SONGS) in Los Angeles proved that this assumption is not generally valid. A simple tube-in-channel theoretical model was previously developed to predict streamwise as well as transverse FEI in a parallel triangular tube array. This predicted that this array geometry was particularly sensitive to streamwise FEI for high mass-damping parameters and small pitch ratios, the conditions in which the SONGS failures occurred. The advantage of this simple modelling approach is that no new empirical data are required for parametric studies of the effects of tube pattern and pitch ratio on FEI. The tube-in-channel model has been extended to in-line square, normal triangular and rotated square tube arrays and the stability of these geometric patterns are analyzed for the effects of varying pitch ratio and the mass-damping parameter. The results are compared with the available experimental data and conclusions are drawn regarding the relative vulnerability of these different tube array geometries to streamwise FEI.

Abstract Power plant component condition monitoring with respect of fatigue and creep-fatigue requires information about the real operational loading of the component. Load monitoring systems are available on the market for that purpose. The retrieved load information can be further processed in terms of fatigue assessment based on realistic loading. Moreover, the combination of load monitoring, derivation of stress-time-histories for sentinel locations, cycle counting and fracture mechanics based assessment paves the way of substantiated determination and optimization of non-destructive testing (NDT) inspection intervals. This approach enables a continuous structural health monitoring of critical components even in the sense of online-monitoring capabilities. This approach provides — in addition to the established fatigue analysis — concrete decisions for the remaining component lifetime and the maximum inspection intervals. This method can also be used if the previous lifetime consumption is unknown. It basically constitutes an advanced damage tolerance approach. The paper describes a newly developed solution which combines the established Fast Fatigue Evaluation (FFE) approach of Framatome with the THERRI capabilities of TÜV NORD.

Small punch test specimens are widely used for a long time as they are simple to produce and requires only a small volume of material. This fact is advantageous especially for high activity materials but also for assessment of operational damage in components materials when component integrity and strength may not be affected. In the same time, no test standard exists and several different specimen types and test procedures have been developed in different place. Thus, to unify this activity, considerable attention has been paid since 2012 to the standardization of small punch test technique within the American Society of Testing and Materials (ASTM). In 2016 a large InterLaboratory Study has been launched within the ASTM subcommittee E10.02 - Behavior and Use of Nuclear Structural Materials, involving 12 laboratories and 6 evaluated structural materials from the nuclear and non-nuclear power plant components. Paper describes the current status of ASTM standardization, results of the InterLaboratory Study, first analysis of the results with respect to some important test parameters, lessons learned and open questions remaining to be solved for the successful completion of the standardization process.

Limit load solutions have been applied to estimate the collapse load of a component made of ductile material. Worldwide maintenance codes for power plants, such as ASME Boiler and Pressure Vessels Code, Section XI, and JSME fitness-for-service code, describe limit load solutions under the assumption of a single flaw. Detected flaws are, however, not always a single flaw, and adjacent flaws due to stress corrosion cracking have been detected in power plants. Thus, development of a limit load solution to estimate the collapse load in the case of multiple flaws remains an issue of structural integrity evaluation. Under the aim of developing a method for evaluating the effect of multiple flaws on collapse load as a part of a limit load solution, fracture tests of flat plates and pipes with multiple flaws were conducted. Although experimental approaches have been attempted to establish the evaluation method, further efforts are required to incorporate the evaluation procedure into a code rule. Effective parameters for considering reduction of collapse load on the basis of test results for specimens with multiple flaws were identified. Test results clearly show a correlation between collapse load and ratios of net-section areas. This correlation leads to the conclusion that distance parameters and flaw length of a smaller flaw determine the existence of an effect on the collapse load by multiple flaws. To investigate the physical sense of the correlation, finite element analysis (FEA) was performed. The FEA results show that strain distributions at the flaw tip under several conditions correspond at the time of maximum load of the fracture tests regardless of the effect of multiple flaws. Also according to the FEA results, the extent of the strain field is linearly proportional to flaw length. These FEA results are consistent with the correlation obtained by the test results.

Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations. Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants. The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.

Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature T oQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T 0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.

A purpose of this study is an improvement of seismic proof construction for power plants to supply power stably at an emergency. At present, the most common type of power generation in Japan is thermal. In particular, coal-fire becomes base load power. A proposal of this study is that boiler structure is applied vibration control. The way is that a damper used viscous fluid is set instead of stopper between the boiler and the support structure. However, inside temperature of the boiler structure is higher than the environment of the general because it uses burning of coal and steam. Therefore, this paper shows that the damper has applicability for the environment of the boiler structure. Then it is necessary for structures to endure earthquake with long duration and long period component in Japan. The 2011 off the Pacific coast of Tohoku Earthquake had the largest energy in history of Japan with them. The damage occurred not only in Tohoku but also in far Tokyo and Osaka. Moreover, it is predicted that large earthquakes with them at the south Pacific coast of Japan occur. This paper shows that the developed damper is effective in earthquake with these characteristics by analyses and component tests.

In 2011, Great East Japan Earthquake that is the largest earthquake ever observed in Japan occurred. The earthquake had large energy, long duration and many aftershocks, and coal-fired thermal power plants were damaged by the earthquake [1]. Boiler structures in coal-fired thermal power plants are generally high-rise structures, and boilers are simply suspended from the top of the support structures in order to allow thermal expansion, so boilers easily vibrate [2]. In order to suppress vibration of boilers during earthquakes, stoppers are generally set between boilers and support structures. The stoppers are made of steel, and dissipate vibration energy by plastic deformation. However aseismic requirements for thermal power plants have been increased as a result of the Great East Japan Earthquake. Thus authors have developed a vibration control damper for coal-fired power plants. The damper is set instead of conventional stopper. Construction of the damper is similar to oil dampers, but inner fluid is viscous fluid. In PVP 2017, the basic performance of the proposed damper was presented [3–5]. In this paper, influence of damper properties on lifetime of the damper was investigated by seismic response analyses. In addition, lifetime of dampers for long period and long duration earthquake waves were investigated by seismic response analyses.

Most nuclear power plants in the US store the spent fuels in independent spent fuel storage casks and these casks are typically placed on concrete pads outside of the fuel handling building. Under plant design basis events, the spent fuel storage casks should maintain stability without tip-over or direct contact with each other. Sliding and rocking of the casks can be determined using nonlinear dynamic analyses under artificial acceleration input motions. Alternatively, approximate equations developed for sliding and rocking of rigid bodies are used as shown in ASCE 4-16. However, these equations consider rocking and sliding as two separate events. Due to the shortcoming of the approximate method, many power plant owners are required to perform extensive nonlinear analyses to ensure cask stability during seismic events. In this study, an independent spent fuel storage cask model is developed and nonlinear dynamic analyses are conducted with seismic input motions that meet the current US Nuclear Regulatory Commission requirements. The analysis results are compared with the approximate method in ASCE 4-16. Based on the comparison, recommendations are made for the use of the approximate approach.

Coal-fired thermal power generation became a very important power source in Japan after Great East Japan Earthquake [1]. Therefore improvement of seismic reliability of the coal-fired thermal power plants is required, because occurrence of very large earthquakes is expected in Japan. Boilers of coal-fired power plants are usually suspended from the upper end of support structures in order to allow thermal expansion of the boilers [2], so boilers easily sway during earthquakes. In order to suppress the vibration, stoppers made of steel are generally installed between boilers and their support structures. Although stoppers made of steel are effective for vibration suppression, further countermeasure for earthquakes having long duration and many aftershocks is required. Authors have developed a vibration control damper for coal-fired power plants. The damper is set instead of conventional stopper. Construction of the damper is similar to oil dampers, but inner fluid is viscous fluid. In PVP2017, the basic performance of the proposed damper was presented [3–5]. In this paper, damper properties were adjusted in order to improve vibration control performance of the damper. Influence of damper properties on the performance was investigated by sensitivity analyses. In addition, influence of dispersion of damper properties was also investigated. Long period and long duration earthquake waves were considered in the analyses.

This study presents strong ground motion simulation methods for the future fragility study of a power plant in Southern Taiwan. The modified stochastic method and empirical Green function method are utilized to synthesize the strong ground motions of specific events. A modified physical random function model of strong ground motions for specific sites and events is presented in this study with verification of sample level. Based on the special models of the source, path, and local site, the random variables of the physical random function of strong ground motions is obtained. The inverse Fourier transform is used to simulate strong ground motions. For the empirical Green function method, the observed site records from small earthquake events occurring around the source area of a large earthquake are collected to simulate the broadband strong ground motion from a large earthquake event. Finally, an application of proposed two simulated methods of this study for simulating the ground motion records of Nishi-Akashi Station at 1995 Kobe earthquake and 2006 Southern Taiwan PingDong earthquake are presented.

One of authors proposed a creep life assessment method for USC boiler pipes that can consider heat-to-heat variations of the creep property of each welded joint, where the creep property of the welded joint is estimated from that of each base metal. In the method, it is assumed that the creep property of each base metal in actual pipes is approximately constant in the thickness direction of the pipes, and test results with small samples cut from base metals at the outer surface of pipes are useful for representing the creep properties of the pipes. In this work, the assumption was examined for five pipes of Grade 91 steel, which had been used for longer than 100,000 h at USC power stations. The microstructure, chemical composition, hardness, void density and remaining creep life were investigated in detail in the thickness direction of the pipes. No difference was observed for these items, except in an area less than about 0.2 mm from the outer surface of the pipes, which means that the assumption in the assessment method is valid except in this area. Therefore, it is suggested that an effective portion of the sample taken from USC boiler pipes to consider heat-to-heat variations of the creep properties of base metals is the material excluding the area less than about 1.0 mm from the outer surface of the pipes.

Grade 91 is widely used for steam pipes and tubes in high temperature boilers of ultra-super critical power plants in Japan. It was reported that creep damage may initiate at the fine grain region within the heat affected zone (HAZ) in welded joints prior to the base metal, so called “Type IV” damage, which causes steam leakage in existing power plants. Therefore, development of creep damage assessment methods is not only an important but also an urgent subject to maintain operation reliability. In order to evaluate creep damage of welded joints based on finite element analyses, creep deformation properties of a base metal, a weld metal and a HAZ have to be obtained from creep tests. However, it is difficult to cut a standard size creep specimen from the HAZ region. Only a miniature size specimen is available from the narrow HAZ region. Therefore, development of creep testing and evaluation technique for miniature size specimens is highly expected. In this study, a miniature tensile type solid bar specimen with 1mm diameter was machined from a base metal, a weld metal and a HAZ of a new and a used Grade 91 welded joints, and creep tests of these miniature specimens were conducted by using a special developed creep testing machine. It was found that creep deformation property is almost identical between the base metal and weld metal, and creep strain rate of the HAZ is much faster than that of these metals in the new welded joint. Relationships between stress and creep strain rates of the base metal and the HAZ in the used welded joint are within scatter bands of those in the new material. On the other hand, creep strain rate of the weld metal in the used welded joint became much faster than that in the new one. Then both the standard size and the miniature size cross weld specimens were machined from the new and the used welded joints and were tested under the same temperature and stress conditions. Rupture time of the miniature cross weld specimen is much shorter than that of the standard size cross weld specimen. The finite element creep analysis of the specimens indicates that higher triaxiality stress yields within the HAZ of the standard size specimen than that of the miniature specimen causing faster creep strain rate in the HAZ of the miniature cross weld specimen.

The small punch (SP) testing technique was applied to five heats of Gr.91 steel, which had been actually used for boiler pipings in different ultra-super critical (USC) power plants for long periods of time, to investigate the applicability of this testing technique to the assessment of heat-to-heat variation of creep property. The SP creep test was carried out at the temperature of 650°C and under the loads of 190, 230, 300 N using a small disk-type specimen ( ϕ 8 × 0.5 mm). The experimental results revealed that the SP creep rupture strength (rupture life) and the deformation rate were different depending on the heat. These differences were qualitatively in good agreement with those observed in the uniaxial creep test. The results obtained in this study indicated that the SP creep testing technique could be a strong tool for the assessment of heat-to-heat variation of in-service boiler pipings.

A computational fluid dynamics (CFD) analysis was performed to predict the transient hydrodynamic loads exerted on the steam generator tubes and the thrust forces on the broken pipe (which is equal to the impingement forces on target structures in the expanding fluid jet path) during a main feed water line break (FWLB) accident at a pressurized water reactor (PWR) power plant. To address a possible severe case of the transient hydrodynamic loads, the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe so that the compressed sub-cooled water would be discharged through the short broken pipe. Thus, a sub-cooled liquid flashing flow through the broken short feed pipe was simulated numerically. Typical results of the prediction were illustrated and discussed. In addition, the present simulation in terms of the transient mass flow rates during the blowdown following the MFLB was compared to other previous calculations. Based on the discussions, the present simulation is considered to be physically plausible and more realistic than other previous predictions.

The EPRI Preventive Maintenance Basis Database (PMBD) has become a standard in the industry to develop, validate, or examine the impact of custom changes to maintenance strategies for common power plant equipment. The PMBD provides failure modes and an indication of frequency of occurrence. Recent feedback from PMBD users has made it clear that including a “Cost Module” to work with PMBD data would be a useful addition to the PMBD program and allow users to view the cost impacts associated with alternate custom maintenance strategies. This paper presents a methodology for the merging of maintenance information extracted from PMBD with cost estimates and additional expert-provided reliability data to estimate a maintenance cost distribution. Additional expert information includes missing data and PM type: monitoring, wear-rate reducing (e.g. oil change), or life-restoring (e.g. refurbishment). The cost distribution is calculated via Monte Carlo simulation and is dependent on the PM plan currently considered. Value-based optimization of the PM plan is performed through Bayesian optimization of the mean PM cost by varying the various PM frequencies. Bayesian optimization iteratively uses Gaussian Process Regression (GPR) to fit a non-parametric meta-model to a noisy objective function. As a part of GPR it is necessary to fit a covariance function that describes the spatial correlation or smoothness of the objective cost function. The meta-model with the covariance function effectively produces a built-in sensitivity analysis for the optimization as well.

Électricité de France (EDF) has developed the Investment Portfolio Optimal Planning (IPOP) software tool [1] to be released with the Integrated Life Cycle Management (ILCM) software tool developed by the Electric Power Research Institute (EPRI) [2]. IPOP is an extremely powerful tool that uses genetic algorithms to provide an optimal strategy for investment in spare components and preventive replacements of multiple components at multiple power plant stations across an entire fleet. A drawback of IPOP is that it requires an extensive amount of user information to run even a single component. In response, Component Optimization Analysis Tools (COATs) was developed to simplify the process of deriving an optimal strategy for purchasing spares and replacements for a single component. This paper describes a two-layer algorithm used in the replacement strategy optimization in COATs. The inner layer consists of a Monte Carlo simulation that estimates the Expected Net Present Value (ENPV) of a given replacement strategy. A strategy consists of: the age of a component at which it needs to be replaced, the age of a component at which a spare should be purchased, years left in the plant at which to skip a scheduled replacement, and the end of life at which the scheduled replacement is skipped; and the years left in the plant at which no more spares are purchased. The Monte Carlo analysis uses these four strategy inputs with component costs, acquisition times, and reliability curves with plant downtime costs to calculate an ENPV for that strategy. The outer layer of the algorithm is an optimization layer that can use either Bayesian optimization or genetic algorithms to maximize the ENPV. These optimization algorithms are routinely available in various software packages and effectively treat the ENPV Monte Carlo as a black box function. An efficiency comparison is given between the two optimization algorithms to demonstrate under which conditions each algorithm out performs the other.

Safety valve closure is employed within power plant piping systems to protect sensitive components from damage due to irregular events causing abrupt pressure variations of the thermal fluid flow. The valve closure creates a sudden obstruction to the flow, generating a pressure wave within the fluid which travels upstream and impacts at the pipe elbows. Such an event is known as steam hammer. This steam hammer pressure wave is capable of producing significant loads and stresses which can disrupt the piping supports as the wave travels throughout the pipe system. Previous studies have shown that the magnitude of these transient loads depend upon the characteristics of the flow, valve closure time, elbow-to-elbow pipe section lengths, the piping system flexibility, and the ‘steepness’ of the pressure transient. The latter effect has been ignored in most steam hammer studies; however, wave steepening has been shown to have a significant effect in cases where the pressure wave travels long distances from the safety valve. This study focuses on Computational Fluid Dynamics (CFD) modeling of rapid valve closure to produce this wave steepening effect and to investigate the significance in terms of transient pipe support loads.