Safety Of Reactor Pressure Vessel Research Articles

Validation of the added mass effect in the narrow gaps of nuclear reactor pressure vessels

Nuclear reactor pressure vessels (RPVs) must be robust to external environmental excitations such as seismic loads. Precise numerical models that predict the dynamic responses to external loads are crucial for evaluating the safety of RPVs. Notably, in RPVs submerged in reactor coolant, it is necessary to examine the added mass effect of the fluid via fluid-structure interaction (FSI) analysis, employing either an analytical approach or a numerical analysis. The present study focuses on modeling and analyzing the added mass effect in the narrow gaps associated with the dense structural layout in the reactor. In this context, this paper validates analytical and numerical models that accurately describe the added mass effect for two cases: (i) simple concentric cylinders with a narrow gap filled with fluid and (ii) a complex small-scale RPV with narrow gaps also filled with fluid. Initially, the analytical model, employing Lagrange's equation of motion and velocity potential, and the numerical model, utilizing FE modeling in acoustic FSI through ANSYS, were validated through experimental modal testing for the added mass effect in the narrow gap between simple concentric cylinders. The analytical model for the concentric cylinders indicates that the added mass diverges to infinity as the gap approaches zero, which should be examined carefully. Through experiments for the concentric cylinders, the analytical and FEM-based added mass effects were validated with a natural frequency error of 3 %. Subsequently, based on the results for the concentric cylinders, a complex numerical model of the small-scale reactor with complex boundary conditions was developed and cross-validated through corresponding experimental modal testing of the small-scale RPV. The added mass model for the narrow gaps in the RPV was validated by substantial agreement of eight representative modes, as demonstrated by a 5 % natural frequency error.

Safety margin determination of the nuclear power plant reactor pressure vessel with taking into account warm pre-stress effect

In case for nuclear power plants long-term service operation over their design life, it is necessary to calculate reactor pressure vessel (RPV) strength and durability acknowledgment (static strength, strength under cyclic and seismic loads, brittle fracture resistance (BFR) include) the as one of the most important NPP structure. Usually, according to the brittle strength assessment, RPV resource is determined, that is, time of its subsequent safe operation. The purpose of this work is assessed BFR RPV at potential emergency accidents (EA) using the Ukrainian warm pre-stress approach. The calculated thermohydrodynamic parameters at EA were used to calculate the stress-strain state of the developed reactor finite element (FE) model. For researching, the most indicative scenarios were selected: where reactor is cooled at a high pressure. In RPV FE model cracks are modeled at the most dangerous places - welds and nozzle. Stress intensity factor (SIF) distribution along crack front and temperature for the most dangerous accidents in terms of BFR are presented in figures. Brittle strength condition is ensured during the nuclear power plants service operation for up to 60 years, which is more than 1.5 times more than the oldest Ukrainian power plant with VVER-1000. For some emergency accidents, warm pre-stress really significantly increased RPV safety margin, but for the most dangerous accidents, the results are the same as without taking into account WPS.

Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

Abstract. The strategy denoted as “in-vessel retention (IVR)” is widely used in severe accident (SA) management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV) integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

Probabilistic fracture mechanics analysis of reactor pressure vessel with underclad and through-clad cracks under pressurised thermal shock transient

Semi-elliptical underclad cracks resulting from the fabrication process of a reactor pressure vessel (RPV) were able to be detected by non-destructive testing method. Meanwhile, after long-term operation under severe conditions, such as high temperature, high pressure, and irradiation, the RPV becomes brittle and susceptible to damage, especially when subjected to pressurised thermal shocks (PTS). Therefore, the probabilistic fracture mechanics (PFM) analysis of RPV with the crack should be applied to evaluate the operation safety. To the best of the authors' knowledge, few studies or computer codes have applied PFM analysis for such cracks. Therefore, this study conducts PFM analysis for cracks by modifying the calculation procedure of FAVOR 12.1 computer code. The results show that during the lifetime of a nuclear power plant, such cracks will not threaten the RPV's safety. Additionally, three methods were proposed to improve FAVOR 12.1's ability to perform PFM analysis for axial through-clad cracking.

Analysis for crack growth regularities in the nozzle-cylinder intersection area of Reactor Pressure Vessel

In this study, linear elastic fracture mechanic analysis for Reactor Pressure Vessel (RPV) of Westinghouse AP1000 is presented to compute stress intensity factor (SIF) of nodes on the crack front. Then crack propagation analysis is carried out under pressure loading condition and cracks are postulated in the in-let nozzle-cylinder intersection area which is also the highest stress concentration area of the RPV. The SIFs of the surface nozzle corner cracks are numerically computed under different loading conditions, and the crack shape evolving regularities under design loading condition are presented, thus providing a useful tool for fracture mechanic design of RPV. The analysis reveals that the 1/4 wall thickness corner crack of in-let nozzle do not endanger the safety of RPV under design condition. The crack has a larger propagation rate at pressure vessel shell than that at the in-let nozzle.

Investigation on the RPV structural behaviors caused by various cooling water levels under severe accident

Abstract In the severe accident (SA) event of core meltdown, the ‘In-vessel retention (IVR)’ mitigation has been certified as an effective strategy for accident management in most advanced nuclear power plants. The traditional concept of IVR was based on the idea that the reactor pressure vessel (RPV) was fully submerged into the cooling water. However, the Fukushima accident showed that the cooling water was insufficient due to the malfunction of water supply system, so the RPV structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels created by the SA condition. In achieving this goal, the structural behaviors were numerically investigated in terms of several field parameters, such as temperature, deformation, stress (strain) and damage. Due to the melting pool on the inside and water cooling on the outside, the high temperature gradient was formed across the wall thickness, so RPV failure was found to be the consequence of creep, plasticity and thermal expansion. According to the requirement on RPV safety during the prescribed time, it must be ensured that the water cooling takes effect in preventing the structural failure under SA condition. Through vigorous investigation, it is found that the RPV safety is secured within the 100 creep h. Furthermore, the structural failure site, time and mode are predicted with consideration of the effect of various water levels. Most importantly, the failure is found to take place at the site aside around the water level.

Comparative Study on Reactor Pressure Vessel Failure Behaviors With Various Geometric Discontinuities Under Severe Accident

The so-called “in-vessel retention (IVR)” is a basic strategy for severe accident (SA) mitigation of some advanced nuclear power plants (NPPs). The IVR strategy is to keep the reactor pressure vessel (RPV) intact under SA like core meltdown condition. During the IVR, the core melt (∼1327 °C) is collected in the lower head (LH) of the RPV, while the external surface of RPV is submerged in the water. Through external cooling of the RPV, the structural integrity is assumed to be maintained within a prescribed period of time. The maximum thermal loading is referred to critical heat flux (CHF) on the inside, while the external surface is considered to perform in the environment of the boiling crisis point (∼130 °C). Due to the high temperature gradients, the failure mechanisms of the RPV is found to span a wide range of structural behaviors across the wall thickness, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Besides CHF, the pressurized core meltdown was another evident threat to the RPV integrity, as indicated in the Fukushima accident on 2011. In illustrating the effects of internal pressures and individual CHF on the failure behaviors, three typical RPVs with geometric discontinuity caused by local material melting were adopted for the comparative study. Through finite-element method (FEM), the RPV structural behaviors were investigated in terms of deformation, stress, plastic strain, creep, and damage. Finally, some important conclusions are summarized in the concluding remark. Such comparative study provides insight and better understanding for the RPV safety margin under the IVR condition.

CFD Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local thermal-hydraulic analysis of a small break loss of coolant accident transient, this paper presents mainly a parametric study that helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel, and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation, which will subsequently assess the associated RPV safety margin factors.

Towards Crack Arrest Testing Using Miniature Specimens

Abstract Crack arrest is an important concept that can be useful to guarantee the safety of reactor pressure vessels. In case of a pressurized thermal shock, a postulated crack can initiate and arrest due to a decrease in driving force combined with an increase in material toughness due to the thermal and neutron embrittlement gradient along the thickness. Due to the size of the specimen and the difficulty in obtaining valid results according to the ASTM E1221-06 standard, “Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, KIa, of Ferritic Steels,” current efforts to develop this technique for irradiated materials are very limited. However, advances in dynamic fracture modeling and elastic-plastic fracture mechanics open the possibility of using miniature crack arrest specimens. We have developed a stiff setup to perform crack arrest tests on precracked Charpy (PCCv) specimens. Different strategies were evaluated to provide enough reduction in driving force to produce arrest. Special attention is given to the starter notch and different starter notch preparations are investigated: precracking, chevron, and brittle weld. Testing configurations were found that guarantee crack arrest with sufficient remaining ligament. The dynamic loading condition is investigated experimentally using strain gauges. Finite element calculations of an arresting crack are also performed in order to provide more insight of the loading after the crack arrest event. Although it is not yet possible to derive the arrest fracture toughness from the PCCv testing, results are encouraging and open new perspectives in the field of crack arrest determination.

Effect of Cladding on the Reactor Pressure Vessel Safety

It is common for the most reactor pressure vessel (RPV) of light-water reactors used today, that the RPV made of ferrite-perlitic steel is cladded with an austenitic layer, which well resists corrosion influence. The cladding is manufactured by welding, the technology (mode of application, number of layers) is different by each manufacturer. This is the layer which comes into contact with the working media first, and is of course in cohesive contact with the RPV material. The aim of this study is to demonstrate the new approach, used for assessing the safety of the reactors today. This approach plays a decisive role in the lifespan calculation of already operating nuclear power stations, including the extension of their service period.

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K I C -reference curve and the J R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.

Development of Debris Coolability Analysis Module in Severe Accident Analysis Code SAMPSON for IMPACT Project

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability.

Sicherheitsanalyse von Reaktordruckbehältern

An evaluation of the safety of nuclear reactor pressure vessels for pressurized water reactors to resist fracture is an important feature in ensuring nuclear reactor safety. This evaluation is performed using fracture mechanics. Basic theory of fracture mechanics The linear, elastic and plastic stress distributions for plane stress and plane strain conditions are derived for the different modes of fracture. The concepts of the linear elastic fracture mechanics and of the crack-opening method are shown. A critical analysis of crack growth and of stress corrosion cracking extends the use of fracture mechanics. Technical application of fracture mechanics on nuclear reactor pressure vessel: Design and calculation of reactor pressure vessels using brittle fracture concepts; estimation of permissible crack sizes for the specification of standards for inspection, hydrostatic pressure test and recurring inspections; evaluation of the operating performance and life expectancy of nuclear reactor pressure vessels including crack growth and stress corrosion cracking. Safety evaluation Using the above derived data, the safety of reactor pressure vessels against fracture is shown by means of: crack size-stress diagrams; brittle fracture diagrams and ductile fracture diagrams; operating stress diagrams.

Fracture mechanics for heavy section steel nuclear pressure vessels

The ability of a pressure vessel steel to resist fracture constitutes an important factor in ensuring nuclear reactor safety. In nuclear reactor pressure vessels some regions (belt-line) are subjected to irradiation damage. Irradiation hardening and embrittlement of mild steels are important in connection with the safety of reactor pressure vessels. Fracture mechanics is considered as providing the best general approach to understanding the brittle fracture phenomenon in reactor pressure vessel steels. Experimental testing of a group of reactor vessel steels utilizing Compact Tension specimens was performed to investigate the application of the fracture mechanics approach to these materials in the irradiated condition. Materials investigated include A533 grade B class 1 base plate and weldment. Compact-tension (CT) fracture mechanics specimens up to two inches in thickness ( 2T- CT ) were irradiated at 550°F to a fluence of 1 to 5 × 10 19 in/cm 2. Tensile and Charpy V-notch specimens from the same materials were also included in the irradiation package to provide correlation data. The fracture toughness of the irradiated pressure vessel material was determined by measurement of K I c , the critical stress intensity factor, which is considered to be a basic material property. The fracture toughness K I c for the unirradiated and post-irradiated conditions and the fatigue crack growth rate as a function of the change in stress intensity ΔK I were utilized to demonstrate the integrity of heavy section nuclear reactor pressure vessels. The application of these parameters in conjunction with the normal coolant system design transients for a typical pressurized water reactor demonstrated that an initial allowable flaw size that could exist at the start of life would not grow to a critical size during the design lifetime of the nuclear reactor pressure vessel (40 years).