Reactor Pressure Vessel Assessment Research Articles

Assessing the master curve reference temperature T0 estimated from the Charpy impact transition curve and consequences on the correlation between Charpy DBTT- and fracture toughness T0-shifts

Integrity assessment of reactor pressure vessels of light water reactors follows well established procedures relying on the surveillance program and defined by codes, standards and nuclear regulatory requirements. These procedures were developed in times where scientific knowledge was limited increasing therefore conservative safety margins to overcome the scientific shortcomings.Taking profit from the development of a number of test techniques and evaluation, it is possible nowadays to propose an alternative interpretation of the test results that provides a complementary information to the standard regulatory approach. In this work, the focus is on the Charpy impact test as it plays a critical role in surveillance programs monitoring reactor vessel degradation induced by neutron irradiation. The development of methodologies allowing experimental determination of fracture toughness from small size specimens triggered alternative approaches of extracting fracture toughness from the Charpy impact test. One of those approaches was recently successfully developed using the instrumented Charpy impact test from which the dynamic and static fracture toughness can be estimated with a reasonably good accuracy. This paper aims to provide an equally performant but simplified procedure allowing estimation of the master curve reference temperature, T0, from the Charpy impact transition curve applied to a large variety of reactor materials and irradiated conditions.

Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor Pressure Vessel Ageing Effect

Ageing of materials stands for the change in their mechanical properties, due to thermodynamic imbalance of the initial condition, and gradually bringing the structure to equilibrium in the presence of sufficient diffusive mobility of the atoms. The current paper presents the main mechanisms of degradation of the mechanical properties of the nuclear power plant equipment metal, as well the models for ageing. The ageing effects and ageing indicators typical for the equipment are defined. The ageing process may cause loss of the appropriate functions. It is important to recognize the ageing effects, as well the ageing indicators. Testing methods are defined for every ageing indicator. The paper considers different national approaches to ageing management of structures, systems, and components (SSCs) at nuclear power plants (NPPs).

Multi-dimensional finite element analyses of OECD lower head failure tests

For severe accident assessment of reactor pressure vessel (RPV), it is important to develop an accurate model that can predict transient thermo-mechanical behavior of the RPV lower head under the given condition. The present study revisits the lower head failure with two- and three-dimensional finite element models. In particular, we aim to give clear insight regarding the effect of the three-dimensionality present in the distribution of the thickness and thermal load of the lower head. For a rigorous validation of the result, both the OLHF-1 and the OLHF-2 tests are considered in this study. The result suggests that the three-dimensional effect is not negligible as far as the failure location is concerned. The non-uniformity of the thickness distribution is found to affect the failure location and time. The thermal load, which may not be axisymmetric in general, has the most significant effect on the failure assessment. We also observe that the creep property can affect the global deformation of the lower head, depending on the applied mechanical load.

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

A probabilistic fracture mechanics (PFM) analysis code PASCAL was developed in Japan for probabilistic integrity assessment of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. To strengthen the practical applications of PFM methodology in Japan, PASCAL has been upgraded to a new version, PASCAL5, which enables PFM analyses of RPVs in both PWRs and boiling water reactors (BWRs) subjected to a broad range of transients, including PTS and normal operational transients. In this paper, the recent improvements in PASCAL5 are described such as the incorporated stress intensity factor solutions and corresponding calculation methods for external surface cracks and embedded cracks near the RPV outer surface. In addition, the analysis conditions and evaluation models recommended for PFM analyses of Japanese RPVs in BWRs are investigated. Finally, PFM analysis examples for core region of a Japanese BWR-type model RPV subjected to two transients (i.e., low-temperature over pressure and heat-up transients) are presented using PASCAL5.

A New Stress Intensity Factor Solution Based on the Response Surface Method for Nozzle Corner Cracks in Nuclear Reactor for Thermal Energy Generation

As considered carbon-free, the use of nuclear energy for thermal energy generation may expand in the future, with the guarantee of safe operation of the nuclear reactor. In a nuclear reactor pressure vessel (RPV), the nozzle area is an important part of the safe operation. It bears internal pressure, axial force, and overall moment, and at the same time bears higher stress than the rest of the vessel. To assess the integrity of the nozzle structure with a crack under combined load, an accurate solution of stress intensity factors (SIF) along the crack front is necessary. To obtain the SIF, this paper proposes a solution method that uses the stress on the crack surface and the response surface method to fit the stress under the framework of the linear superposition technique. This method is the first choice to determine a series of influence coefficients under unit pressure load. Then, one can estimate the SIFs by superposition method for an arbitrary stress distribution resulted from combined loads. The proposed solution is verified for a typical RPV with cracks under internal pressure, axial force, and global bending moment. The results reveal that the proposed solution is in good agreement with the existing solutions under internal pressure. Therefore, it can be obtained that this solution can be effectively used for the structural integrity assessment of RPV with nozzle corner cracks.

A method for directly measuring fracture toughness and determining reference temperature for RPV steels by Charpy impact test

Accurately realizing the use of Charpy impact test to measure fracture toughness and determine reference temperature of low alloyed steel with ferritic microstructure has important practical engineering significance for the structural integrity assessment of reactor pressure vessels. In this paper, a methodology for measuring fracture toughness and determining reference temperature for RPV steels by Charpy impact test was proposed. The method referred to the ASTM E1820 and E1921 standard. By analyzing the physical meaning of the Charpy impact specimen’s fracture, load-time curve and size parameters, the fracture toughness of the material was directly calculated according to the linear elastic fracture mechanics. The reference temperature was calculated using the multi-temperature procedure formula. The reference temperature and fracture toughness were corrected according to two factors, such as the difference in fracture toughness caused by the difference in the loading rate of the stress intensity factor and the specimen shape and size between Charpy impact test and compact tension test. And the validity of the fracture toughness and reference temperature was evaluated. The corrected two are basically consistent with the data obtained by the compact tension test in accordance with ASTM E1820 and E1921 standards. This method may be used to solve the problem that the nuclear power plant needs to use the impact test data to evaluate the fracture toughness and reference temperature of the reactor pressure vessel in the structural integrity assessment of the reactor pressure vessel operation and life extension process.

Effect of Coolant Water Temperature of Emergency Core Cooling System on Failure Probability of Reactor Pressure Vessel

Abstract Structural integrity assessment of reactor pressure vessel (RPV) is important for the safe operation of nuclear power plant. For an RPV in a pressurized water reactor (PWR), pressurized thermal shock (PTS) resulted from rapid coolant water injection due to a loss-of-coolant accident is an issue of particular concern. The coolant water temperature in the emergency core cooling system (ECCS) can influence the integrity of RPV subjected to PTS events; thus, this paper is focused on investigating the effect of coolant water temperature of ECCS on failure probability of an RPV. First, thermal-hydraulic (TH) analyses were conducted for a Japanese PWR model plant by using RELAP5, and different coolant water temperatures in ECCS were considered to investigate the effect of coolant water temperature on TH behaviors during a PTS event. Using the TH analysis results, probabilistic fracture mechanics (PFM) analyses were performed for the RPV of the Japanese model plant. Based on the PFM analysis results, the effect of coolant water temperature on failure probability of the RPV was quantified.

Ductile fracture properties of 16MND5 bainitic forging steel under different in-plane and out-of-plane constraint conditions: Experiments and predictions

Bridging the gap between small-scale laboratory specimens and engineering full size engineering structures plays a critical role in integrity assessment of reactor pressure vessels (RPVs). In the present work, in-plane and out-of-plane constraint effects are both considered to determine ductile fracture properties under various low constraint conditions. The fracture behavior of 16MND5, a bainitic forging steel was investigated using clamped single edge notched tension (SENT) specimens with different crack sizes, specimen thicknesses and span lengths. Various fracture properties, including R-curves (J-integral, CTOD) and initiation fracture toughness (JSZW, J0.2, CSZW and C0.2) were determined. The results showed that the fracture properties are highly dependent on the in-plane (crack depth) and out-of-plane (specimen thickness) constraint conditions. The lower in-plane and out-of-plane constraint levels will introduce higher fracture properties. To further understand and quantify the impact of in-plane and out-of-plane constraint effects, the finite element method (FEM) and fracture surface analysis were conducted. Constraint dependent R-curves and initiation fracture toughness are developed for 16MND5 steel using a newly developed three-parameter fracture mechanics methodology. In particular, two independent constraint parameters, one for in-plane and one for out-of-plane, were used to quantify the 3D constraint effect. Finally, the present method was used to predict fracture properties for other specimens or low-constraint RPV cracked structures with different in-plane and out-of-plane constraint conditions.

Application of Flaw Updating Process on Probabilistic Integrity Analysis for a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks

The structural integrity of a reactor pressure vessel (RPV) is a crucial issue for an operating nuclear power plant, especially in the beltline region, which suffers the highest neutron irradiation. Owing to its capability of considering parameters based on statistical distributions and provision of objective risk-informed results, the probabilistic fracture mechanics (PFM) method is widely used in evaluating the structural integrity of RPVs. However, the flaw characteristics used for PFM analysis are mainly derived from particular vessel inspection information such as from the Pressure Vessel Research User Facility and Shoreham vessels, which may not be able to truly represent the vessel-specific condition of an analyzed RPV. In this work, the Bayesian inference, which combines prior flaw data with new inspection results as well as uncertainties, is used to develop posterior vessel-specific flaw distributions. Then, the updated flaw model is used for PFM analysis to investigate the effects on the fracture probability assessment of RPVs subjected to pressurized thermal shocks (PTSs). Considering the updated flaws based on the inspection data, the PFM analysis result could be more realistic to predict the fracture risks of RPVs during operation.

Plasticity Correction on Stress Intensity Factor Evaluation for Underclad Cracks in Reactor Pressure Vessels

Abstract Structural integrity assessment of reactor pressure vessels (RPVs) is essential for the safe operation of nuclear power plants. For RPVs in pressurized water reactors (PWRs), the assessment should be performed by considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. To assess the structural integrity of an RPV, a traditional method is usually employed by comparing fracture toughness of the RPV material with the stress intensity factor (KI) of a crack postulated near the RPV inner surface. When an underclad crack (i.e., a crack beneath the cladding of an RPV) is postulated, KI of this crack can be increased owing to the plasticity effect of cladding. This is because the yield stress of cladding is lower than that of base metal and the cladding may yield earlier than base metal. In this paper, detailed three-dimensional (3D) finite element analyses (FEAs) were performed in consideration of the plasticity effect of cladding for underclad cracks postulated in Japanese RPVs. Based on the 3D FEA results, a plasticity correction method was proposed on KI calculations of underclad cracks. In addition, the effects of RPV geometries and loading conditions were investigated using the proposed plasticity correction method. Moreover, the applicability of the proposed method to the case which considers the hardening effect of materials after neutron irradiation was also investigated. All of these results indicate that the proposed plasticity correction method can be used for KI calculations of underclad cracks and is applicable to structural integrity assessment of Japanese RPVs containing underclad cracks.

Engineering critical assessment of RPV with nozzle corner cracks under pressurized thermal shocks

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

Improvements on Evaluation Functions of a Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessels

Abstract In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.

ANALYSIS OF 3D SEMI-ELLIPTICAL CRACK ON REACTOR PRESSURE VESSEL WALL WITH LOAD STRESS AND CRACK RATIO

Reactor Pressure Vessel (RPV) wall is an important component in the Nuclear Power Plant (NPP). During reactor operation, RPV is subjected to high temperature, pressure, and neutron exposure. This condition could lead to RPV structure failure. In order to assure the integrity of RPV during the reactor lifetime, it is mandatory to perform a structural integrity assessment of RPV by evaluating postulated crack in RPV. In the previous study, the crack has evaluated in 2-D. However, 3-D analysis of semi-elliptic crack shape in the surface of the thick plate for RPV wall using SA 508 Steel is yet to be analyzed. The objective of this study is to analyze and modeling the evaluation in variation crack ratio with some load stress in 3-D. The Stress Intensity Factor (SIF) and J-integral are used as crack parameter. The J-Integral were calculated using MSC MARC MENTAT based on Finite Element Method (FEM) for obtaining the SIF value. The inputs are a crack ratio, load stress, material property, and geometry. The modeling of SIF value and goodness of fit are using MINITAB. The fracture condition could be predicted in comparison to the SIF value and fracture toughness. For the load stress 70 MPa and 80 MPa, with a crack ratio 0.25, 0.33 and 0.5, the material on RPV wall will in fracture condition.Keywords: Semi elliptic surface crack, 3-dimension, reactor pressure vessel, elastic-plastic fracture mechanics, J-integral

Influence Factors of Fracture Mechanics Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock

When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

A modified correlation between KJIC and Charpy V-notch impact energy of Chinese SA508-III steel at the upper shelf

The fracture toughness plays a significant role in the structural integrity assessment of reactor pressure vessels (RPVs) in service temperature. The Charpy V-notch (CVN) impact test is used to estimate fracture toughness (KIC or KJIC) indirectly since universal fracture toughness tests are costly, sophisticated and frequently invalid. In this study, a modified correlation which based on the typical model of KJIC-CVN at the upper shelf was established for Chinese SA508-III steel. Thereinto, the effect of test temperature (T) was directly considered in the correlation. To assess the accuracy of fracture toughness when calculating from the value of Charpy-V notch impact energy by using the modified correlation, both the Charpy-V notch impact tests and fracture toughness tests for Chinese SA508-III steel were conducted at different temperatures (100 °C, 150 °C, 200 °C, 250 °C and 320 °C). The results showed that the modified correlation exhibited the high precision for estimating fracture toughness of Chinese SA508-III steel and the relative error for tested and estimated results is within 8%, which is lower than that of other correlations.

Constraint Assessment for Specimens Tested Under Uniaxial and Biaxial Loading Conditions

In structural integrity analysis of reactor pressure vessels (RPVs), a postulated shallow crack is subjected to biaxial far-field stresses. However, the fracture toughness Kc or Jc, which is an important material property for the structural integrity assessment of RPVs, is usually obtained from testing deeply cracked compact tension (C(T)) or single-edged bending (SE(B)) specimens under uniaxial loading. Thus, the fracture toughness data do not reflect the biaxial loading state that cracks in a RPV are subjected to. Cruciform bending specimen was therefore developed to simulate the biaxial stress state. In this paper, a series of finite element (FE) simulations of the cruciform specimens containing different crack geometries and of different material properties are conducted. The crack tip stress fields are analyzed, and the constraint is investigated using the J–A2 theory. The results show that the biaxial effect is material property dependent which could be useful for the optimization of the test method and the better design of cruciform specimens. The trends about the biaxial loading effect revealed in this study would also be helpful in estimating the safe operating life of RPVs.

Ultimate bearing capacity analysis of a reactor pressure vessel subjected to pressurized thermal shock with XFEM

Abstract Under pressurized thermal shock (PTS), the structure integrity of reactor pressure vessel (RPV) meets a potential challenge. The vicinity of inlet nozzle is the weak link and hazardous area of the vessel. Once crack initiation occurs here, the stress concentration around the crack tip would lead to plastic deformation, crack propagation, and even the failure of the whole structure. Based on ABAQUS, the beltline region around the inlet nozzle was modeled to calculate the response of the transient temperature field and stress strain field. In order to demonstrate the ultimate bearing capacity of the RPV with the critical crack sizes, the dynamic process of crack propagation was simulated with the XFEM. Through the quantitative analysis of the Mises stress and circumferential stress through the wall thickness, the crack propagation paths in different directions were confirmed. The above analysis and calculation provide an important reference for the safety and reliability assessment of RPV.

Blunt defect assessment in the framework of the failure assessment diagram

In order to reduce over-conservatism in fitness-for-service assessment procedures, experimental evidence and recent analytical developments recognise the importance of considering the actual shape of non-sharp flaws and/or the real geometric constraint conditions at the crack tip. This paper addresses the effect of blunt defects on the structural integrity assessment of reactor pressure vessel (RPV) and pipeline steels. Parametric studies for compact tension specimens with various notch root radii are performed using finite element analysis. The notch fracture toughness, the resistance to the onset of ductile cracking and the J-integral, quantifying the notch driving force, are evaluated. A stress-modified fracture strain model is used as a virtual testing method. The results are analyzed in the framework of the failure assessment diagram (FAD), showing that the existing shape of the FAD is also suitable for assessments of blunt defects and how the concepts introduced can be used to reduce the conservatism in defect assessment, define margins on failure and indicate when plastic collapse is the dominant failure mechanism.

On Issue of Comparison of the Unified Curve and Master Curve Methods and Application for RPV Structural Integrity Assessment

Application of two engineering methods, the Master Curve and the Unified Curve, has been analyzed for structural integrity assessment of reactor pressure vessels. These methods have been compared using the fracture toughness data base consisting of 44 sets for ferritic steels and their welds with various degrees of embrittlement. As the comparison tests of the Master Curve and the Unified Curve, three statistical parameters have been used. Treatment of the fracture toughness data base has shown the advantage of the Unified Curve over the Master Curve when using all three parameters. Possible reasons of illusory advantage of the Master Curve over the Unified Curve that is revealed sometime when comparing these methods have been analyzed, and the main requirements have been formulated and substantiated for fully objective comparison of the Master Curve and the Unified Curve. It has also been explained with the Prometey brittle fracture model and the experimental data on the temperature dependence of thermal part of yield stress why the K JC (T) dependencies for different ferritic steels with small degrees of embrittlement may be approximated by the same exponential function as taken for the Master Curve. The correlation has been found between the temperature dependence of thermal part of yield stress and the Master Curve.