Fracture Analysis of Surface-Breaking Shallow Flaw Behaviour in the NESC-IV Bend Beam Tests

2004 ◽  
Author(s):  
N. Taylor ◽  
I. Sattari-Far ◽  
D. Siegele ◽  
I. Varfolomeyev ◽  
L. Stumpfrock
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Biaxial Loading ◽  
Production Quality ◽  
Standard Test ◽  
Reactor Pressure Vessel ◽  
Weld Material ◽  
Post Test ◽  
Analytical Program ◽  
Reactor Pressure

NESC-IV is an experimental/analytical program to develop validated analysis methods for transferring fracture toughness data generated on standard test specimens to shallow flaws in reactor pressure vessel welds subject to biaxial loading in the lower-transition temperature region. The testing program exploited material from a production-quality reactor pressure vessel. In Part A six clad cruciform specimens containing shallow surface-breaking flaws located in weld material were successfully tested. Post-test fracture mechanics analyses have been conducted by several organizations to determine fracture mechanics parameters such as K, J, Q and T at the flaws. These have been used to interpret the results with respect to the Master Curve, giving particular attention to constraint aspects.

Overview of NESC-IV Cruciform Specimen Test Results

2002 ◽  
Author(s):  
N. Taylor ◽  
R. Bass
Keyword(s):  
Pressure Vessel ◽  
Best Practice ◽  
Biaxial Loading ◽  
Production Quality ◽  
Standard Test ◽  
Reactor Pressure Vessel ◽  
Test Analysis ◽  
Post Test ◽  
Specimen Test ◽  
Reactor Pressure

NESC-IV is an experimental/analytical program to develop validated analysis methods for transferring fracture toughness data generated on standard test specimens to shallow flaws in reactor pressure vessel welds subject to biaxial loading in the lower-transition temperature region. It is the fourth major project of the Network for Evaluating Structural Components (NESC). The testing program has exploited material removed from a production-quality reactor pressure vessel. In Part A six clad cruciform specimens containing shallow surface-breaking flaws located in weld material were successfully tested. For Part B a further four beam tests were performed using an innovative test piece design with a simulated embedded flaw. Post-test analysis is now in progress. The implications for current best-practice procedures for evaluation of RPV shallow are also being considered.

Use of Master Curve Technology for Assessing Shallow Flaws in a Reactor Pressure Vessel Material

2008 ◽  
Vol 130 (3) ◽  
Author(s):  
N. Taylor ◽  
P. Minnebo ◽  
B. R. Bass ◽  
D. Siegele ◽  
K. Wallin ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Master Curve ◽  
Biaxial Loading ◽  
Production Quality ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Test Pieces ◽  
Fracture Tests ◽  
Vessel Material ◽  
Reactor Pressure

In the NESC-IV project, an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope, an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The master curve analysis of these data is reported and its application to the assessment of the project feature tests on large beam test pieces is discussed.

Use of Master Curve Technology for Assessing Shallow Flaws in a Reactor Pressure Vessel Material

2006 ◽  
Author(s):  
N. Taylor ◽  
P. Minnebo ◽  
R. B. Bass ◽  
D. Siegele ◽  
K. Wallin ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Master Curve ◽  
Biaxial Loading ◽  
Production Quality ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Test Pieces ◽  
Fracture Tests ◽  
Vessel Material ◽  
Reactor Pressure

In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

scholarly journals Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

2020 ◽  
Vol 7 (3) ◽  
pp. 19-00573-19-00573
Author(s):  
Kai LU ◽  
Jinya KATSUYAMA ◽  
Yinsheng LI ◽  
Yuhei MIYAMOTO ◽  
Takatoshi HIROTA ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
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 the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Fracture Mechanics Assessment of Reactor Pressure Vessel Irradiated Structural Steel for Short Column Type Supports and Neutron Shield Tank

2020 ◽  
Author(s):  
Alexandria M. Carolan ◽  
J. Brian Hall ◽  
Stephen K. Longwell ◽  
F. Arzu Alpan ◽  
Gregory M. Imbrogno ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Structural Steel ◽  
Pressure Vessel ◽  
Radiation Effects ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Short Column ◽  
License Renewal ◽  
Reactor Pressure

Abstract As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.

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