Assessment of a Stress-Free Temperature Model for Residual Stresses in Surface Cladding of a Reactor Pressure Vessel

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Thermal Expansion ◽  
Residual Stresses ◽  
Base Metal ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Normal Operation ◽  
Inner Surface ◽  
Reactor Pressure

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature (Ts-free), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

Residual Stresses in Clad Nuclear Reactor Pressure Vessel Steels: Prediction, Measurement and Reconstruction

2015 ◽  
Author(s):  
Karim Serasli ◽  
Harry Coules ◽  
David Smith
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Residual Stresses ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Finite Element Simulations ◽  
Reactor Pressure Vessel ◽  
Mapping Technique ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Most residual stress measurement methods are limited in terms of their stress and spatial resolution, number of stress tensor components measured and measurement uncertainty. In contrast, finite element simulations of welding processes provide full field distributions of residual stresses, with results dependent on the quality of the input conditions. Measurements and predictions are often not the same, and the true residual stress state is difficult to determine. In this paper both measurements and predictions of residual stresses, created in clad nuclear reactor pressure vessel steels, are made. The measurements are then used as input to a residual stress mapping technique provided within a finite element analysis. The technique is applied iteratively to converge to a balanced solution which is not necessarily unique. However, the technique aids the identification of locations for additional measurements. This is illustrated in the paper. The outcomes from the additional measurements permit more realistic and reliable estimates of the true residual state to be made. The outcomes are compared with the finite element simulations of the welding process and used to determine whether there is a need for additional input to the simulations.

Welding Residual Stress and Multiple Flaw Evaluation for Reactor Pressure Vessel Head Replacement Welds With Alloy 52

2009 ◽  
Author(s):  
T. Zhang ◽  
F. W. Brust ◽  
G. Wilkowski ◽  
D. L. Rudland ◽  
A. Csontos
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Limit Analysis ◽  
Pressure Vessel ◽  
Hot Cracking ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Multiple Cracks ◽  
Head Penetration ◽  
Reactor Pressure

Small indications were found in one replacement reactor pressure vessel head (RPVH) mock-up being fabricated from Alloy 690 material and compatible weld metals, Alloy 52/152. The mockups were non-destructively examined and the lowest number of cracks found was five and the highest number was 22. There are numerous indications with some of them quite long (50 mm) in length. The source of these weld fabrication cracks is unknown. However, from experience with other difficult to weld materials, the source can range from slag inclusions in the weld metal to hot cracking during the weld deposition process. Hot cracking includes solidification cracking (weld), liquation cracking (HAZ), and ductility dip cracking (DDC). The indications were mostly circumferential in orientation (with respect to the nozzle axis) but some were axial. This paper includes two parts. The first part includes the welding residual stress analysis of RPVH using Alloy 52/152 metal and provides comparison with similar Alloy 82/182 welds. Alloy 82/182 was the material used in the original dissimilar metal welds in these heads. Primary Water Stress Corrosion Cracking (PWSCC) can occur in the primary coolant system when the welds are exposed to water, tensile stress, and temperature (usually higher than 250 C). PWSCC rates are higher in Alloy 82/182 material due to its lower chromium content compared with the replacement alloy. The results for both center hole (0-degree) and side hill (53-degree) nozzles will be discussed. The second part deals with assessment of multiple small cracks in the reactor pressure vessel head penetration nozzles. The finite element alternating method (FEAM) was used for calculating stress intensity factors for cases where multiple cracks exist. More than twenty cracks, which were inserted based on field measurements, are considered in the analyses for both center hole and side hill nozzles. It is observed that the overall stress trends are similar to those without adding cracks. However, cracks introduce more local stress fluctuations. The magnitude of the local fluctuation can be around 100MPa. Limit analysis was also conducted. A new finite element model with a voided-out weld region was used to simulate loss of structural capacity due to multiple flaws. The voided out volume effects on the structural integrity and future performance of RPVH were examined. Discussions based on weld residual stress, multiple flaw analysis and limit analysis conclude the paper.

ENPOWER: Repair Welds and Residual Stresses in Clad Plates

2005 ◽  
Author(s):  
Hieronymus Hein ◽  
Bruce Brown ◽  
Didier Lawrjaniec ◽  
Carsten Ohms ◽  
Chris Truman ◽  
...  
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Numerical Simulations ◽  
European Commission ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Stress Measurements ◽  
Reactor Pressure

One of the tasks of the European Commission sponsored project ENPOWER was to manufacture repair welds on clad plate specimens simulating the inner wall of a Reactor Pressure Vessel (RPV) and to establish their structural integrity. The paper summarizes the main results from the repair welds carried out on clad plates with an anticipated sub-clad defect including the results from various residual stress measurements and from numerical welding simulations as well as from some fracture mechanical calculations. The results are discussed with respect to support the repair weld optimization in particular by minimizing the residual stresses. Moreover, the application ranges and capabilities of numerical simulations for this kind of weld processes are discussed.

Prediction of Residual Stress Distribution for RPV CRDM Penetration Nozzles

2012 ◽  
Author(s):  
Hong-Yeol Bae ◽  
Yun-Jae Kim ◽  
Ju-Hee Kim ◽  
Sung-Ho Lee ◽  
Kyoungsoo Lee
Keyword(s):  
Residual Stress ◽  
Yield Strength ◽  
Residual Stresses ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Alloy 600 ◽  
Tube Radius ◽  
Pressurized Water ◽  
Tube Geometry ◽  
Reactor Pressure

In nuclear power plants, RPV (Reactor Pressure Vessel) upper head CRDM (Control Rod Drive Mechanism) penetration tubes has been fabricated J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM tubes and their associated welds have increased significantly. The cracking mechanism has been attributed to PWSCC (Pressurized Water Stress Corrosion Cracking) and has been shown to be driven by welding residual stresses and operational stresses in the weld region. Weld induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operating. In this point, we have been conducting detailed welding simulation analyses for Korea Nuclear Reactor Pressure Vessel to predict the magnitude of weld residual stresses in penetration tubes. In the present work, the FE (Finite Element) simulations were conducted to investigate the effects of tube geometry (location and ro/t) and material properties on the residual stresses in the J-groove weld for a different location of CRDM tubes. The variables of tube location included three (center-hole, intermediate and steepest side hill tube) inclination angles (Ψ). And this comparison was performed for different tube geometry (ro/t = 2, 3, 4), different yield strength (σo) of tube. In CRDM tube, when increases in tube inclination angle (Ψ), axial residual stress are gradually increased, but hoop residual stresses are decreased at the nearby weld root. In effect of tube radius and thickness, when the thickness of CRDM tubes increases the residual stresses are gradually decreased at the inner surface of tube. And there is no effect of CRDM tube radius (ro). In effect of plastic properties of Alloy 600 material in CRDM, when yield strength of the tube increases the axial residual stresses decreases but hoop residual stress increases.

Fast Neutron Irradiation Optimization of Thorium-Fueled SCWR Reactor Pressure Vessel

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Petra Pónya ◽  
Gyula Csom ◽  
Sándor Fehér
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Flow Direction ◽  
Reactor Pressure Vessel ◽  
Fast Neutron Flux ◽  
Inner Surface ◽  
Fast Neutron Irradiation ◽  
Reactor Pressure

Abstract Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at ∼500 °C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

2009 ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Root Region ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

Structural Integrity of Penetration Nozzle of Reactor Pressure Vessel Head of PWR Plant

2008 ◽  
Author(s):  
Kiminobu Hojo ◽  
Naoki Ogawa ◽  
Yoichi Iwamoto ◽  
Kazutoshi Ohoto ◽  
Seiji Asada ◽  
...  
Keyword(s):  
Residual Stress ◽  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Stress Field ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Complex Structure ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

A reactor pressure vessel (RPV) head of PWR has penetration holes for the CRDM nozzles, which are connected with the vessel head by J-shaped welds. It is well-known that there is high residual stress field in vicinity of the J-shaped weld and this has potentiality of PWSCC degradation. For assuring stress integrity of welding part of the penetration nozzle of the RPV, it is necessary to evaluate precise residual stress and stress intensity factor based on the stress field. To calculate stress intensity factor K, the most acceptable procedure is numerical analysis, but the penetration nozzle is very complex structure and such a direct procedure takes a lot of time. This paper describes applicability of simplified K calculation method from handbooks by comparing with K values from finite element analysis, especially mentioning crack modeling. According to the verified K values in this paper, fatigue crack extension analysis and brittle fracture evaluation by operation load were performed for initial crack due to PWSCC and finally structural integrity of the penetration nozzle of RPV head was confirmed.

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