Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

2011 ◽  
Author(s):  
Shengjun Sean Yin ◽  
Gary L. Stevens ◽  
B. Richard Bass ◽  
Mark T. Kirk
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Pressurized Water Reactor ◽  
Outlet Nozzle ◽  
Corner Crack ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Reactor Pressure

This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle; 3. PWR inlet nozzle; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: • To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; • To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; • To assess the significance of attached piping loads on the stresses in the nozzle corner region; and • To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

Stress Intensity Factor Influence Coefficients for External Surface Flaws in Boiling Water Reactor Pressure Vessels

2009 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
External Surface ◽  
Pressure Vessels ◽  
Water Reactor ◽  
Influence Coefficients ◽  
Surface Flaws ◽  
Reactor Pressure

Over the service life of a nuclear power plant, the Boiling Water Reactor (BWR) may undergo many cool-down and heat-up thermal-hydraulic transients associated with, for example, scheduled refueling outages and other normal operational transients, or even possible overcooling transients. These thermal-hydraulic events can act on postulated surface flaws in BWRs and therefore impose potential risk on the structure integrity of Reactor Pressure Vessels (RPVs). Internal surface flaws are of interest for the BWRs under overcooling accidental scenarios, while external surface flaws are more vulnerable when the BWRs are subjected to heat-up transients. Stress Intensity Factor Influence Coefficient (SIFIC) databases for application to linear elastic fracture mechanics analyses of BWR pressure vessels which typically have an internal radius to wall thickness ratio (Ri/t) between 15 and 20 were developed for external surface breaking flaws. This paper presents three types of surface flaws necessary in fracture analyses of BWRs: (1) finite-length external surface flaws with aspect ratio of 2, 6, and 10. (2) infinite-length axial external surface flaws; and (3) 360° circumferential external surface flaws. These influence coefficients have been implemented and validated in the FAVOR fracture mechanics code developed at Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Although these SIFIC databases were developed in application to RPVs subjected to thermal-hydraulic transients, they could also be applied to RPVs under other general loading conditions.

Supplemental Stress and Fracture Mechanics Analyses of Pressurized Water Reactor Pressure Vessel Nozzles

2012 ◽  
Author(s):  
Matthew Walter ◽  
Shengjun Yin ◽  
Gary L. Stevens ◽  
Daniel Sommerville ◽  
Nathan Palm ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Additional Stress ◽  
Corner Crack ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Reactor Pressure

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: • To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). • To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. • To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, “Fracture Toughness Requirements,” and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

Research on the Fracture Mechanics Analysis Method for Reactor Pressure Vessel Nozzle Corner Flaw

2017 ◽  
Author(s):  
Shen Rui ◽  
Cao Ming ◽  
He Yinbiao ◽  
Tao Hongxin
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Solution Method ◽  
Reactor Pressure Vessel ◽  
Design Code ◽  
Outlet Nozzle ◽  
Reactor Pressure

This paper has discussed the stress intensity factor solution method of most popular used nuclear equipment design code and published papers. A series of inlet and outlet nozzle of reactor pressure vessel 3-D FEA fracture mechanics models with different size of corner flaw are created by ABQUAS software. Moreover, the crack front has been specially processed by ZENRCAK software. By compare the stress intensity factor solutions of FEA method and the solutions of influence function method for a 1/4 infinite symmetry plate, the influence functions for PWR reactor pressure vessel inlet and outlet nozzle corner flaw solution are obtained.

An Investigation of the Stress Intensity Factor Along a Corner Crack in Pressurizer Vent Nozzle Penetration Weld

2009 ◽  
Author(s):  
Sang-Min Lee ◽  
Jeong-Soon Park ◽  
Jin-Su Kim ◽  
Young-Hwan Choi ◽  
Hae-Dong Chung
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Plastic Zone ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Corner Crack ◽  
Elastic Plastic ◽  
Material Fracture

Elastic-plastic fracture mechanics as well as linear-elastic fracture mechanics may be applied to evaluate a flaw in ferritic low alloy steel components for operating conditions when the material fracture resistance is controlled by upper shelf toughness behavior. In this paper, the distribution of the stress intensity factor along a corner crack using elastic-plastic fracture mechanics technique is investigated to assess the effect of a structural factor on mechanical loads in pressurizer vent nozzle penetration weld. For this purpose, the stress intensity factor and plastic zone correction of a corner crack are calculated under internal pressure, thermal stress and residual stress in accordance with Electric Power Research Institute (EPRI) equation and Irwin’s approach, respectively. The resulting stress intensity factor and plastic zone correction were compared with those obtained from Structural Integrity Associates (SIA) and Kinectrics, and were observed to be good agreement with Kinectrics results.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Effect of Cladding Residual Stress Modeling Technique on Shallow Flaw Stress Intensity Factor in a Reactor Pressure Vessel

2015 ◽  
Author(s):  
Joshua Kusnick ◽  
Mark Kirk ◽  
B. Richard Bass ◽  
Paul Williams ◽  
Terry Dickson
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Fracture Mechanics ◽  
Intensity Factor ◽  
Residual Stresses ◽  
Room Temperature ◽  
Pressure Vessels ◽  
Temperature Model ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

Prior probabilistic fracture mechanics (PFM) analysis of reactor pressure vessels (RPVs) subjected to normal cool-down transients has shown that shallow, internal surface-breaking flaws dominate the RPV failure probability. This outcome is caused by the additional crack driving force generated near the clad interface due to the mismatch in coefficient of thermal expansion (CTE) between the cladding and base material, which elevates the thermally induced stresses. The CTE contribution decreases rapidly away from the cladding, making this effect negligible for deeper flaws. The probabilistic fracture mechanics code FAVOR (Fracture Analysis of Vessels, Oak Ridge) uses a stress-free temperature model to account for residual stresses in the RPV wall due to the cladding application process. This paper uses finite element analysis to compare the stresses and stress intensity factor during a cool-down transient for two cases: (1) the existing stress-free temperature model adopted for use in FAVOR, and (2) directly applied RPV residual stresses obtained from empirical measurements made at room temperature. It was found that for a linear elastic fracture mechanics analysis, the application of measured room temperature stresses resulted in a 10% decrease in the peak stress intensity factor during a cool-down transient as compared to the stress-free temperature model.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

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