Reactor Pressure Vessel Life Management at French PWR Plants

2002 ◽  
Author(s):  
Anne Mermillod ◽  
Georges Bezdikian ◽  
Serge Rostain
Keyword(s):  
Electricity Generation ◽  
Safety Margin ◽  
Ductile Failure ◽  
Mechanical Analysis ◽  
Reactor Vessel ◽  
Experimental Program ◽  
Surveillance Programme ◽  
Input Parameters ◽  
Reactor Pressure

The capability of nuclear electricity generation by French Utility’s EDF (Electricite´ de France) is 75% of global generation, 58 PWR units are in operation, 34 units three-loop, 20 units four-loop (1300MWe) and 4 units four-loop (1450MWe). The French Utility methodology to justified reactor vessel assessment on specific plants is to verify, for a known defect (real or hypothetic), safety margin to non ductile failure for each transient and for case of loading. The demonstration must be done for each instant of the reactor vessel life, and must include material embrittlement caused by irradiation degradation. This paper will describe EDF methodology aging assessment for at least 40 years, comparing: • The boundary RTNDT (Reference Nil Ductility Transition Temperature) support by reactor vessel. • The estimated RTNDT at 40 years. Estimated RTNDT for 40 years is determined with a prediction formula that needs following input parameters: • The initial RTNDT during the construction. • Fluence assessment received by reactor vessel during 40 years in operation. • Chemistry material composition. The Boundary RTNDT supported by reactor vessel is determined with a mechanical analysis that needs the following input parameters: • Defect characteristics. • Transient definition. The determination of each parameter will be described, and an experimental program to verify the conservatism of methodology will be presented (Irradiation Surveillance Programme).

Mechanical Analysis of Underclad Cracks in Reactor Pressure Vessel Nozzles

2003 ◽  
Author(s):  
Nicolas Verdiere ◽  
Henriette Churier-Bossennec ◽  
Franc¸ois Champigny
Keyword(s):  
Mechanical Analysis ◽  
Reactor Vessel ◽  
Core Shell ◽  
Straight Section ◽  
The Core ◽  
Inner Radius ◽  
Service Inspection ◽  
Reactor Pressure ◽  
Circular Cracks ◽  
Fast Fracture

During the in-service inspection in 1999 in Tricastin 1 Nuclear Plant, underclad cracks had been discovered in the core shell. Following specific justification of these defects, other parts of the reactor vessel also sensitive to this kind of defects have been re-examined, and particularly RPV nozzles, which are classified as “concerned by fast fracture”. Recent assessments with large defects, superior to maximum realistic ones, have been conducted in different sections in the reactor vessel inlet and outlet nozzles: circular cracks of 10 mm in the inner radius, and 20 mm depth in the straight section. Theses assessments highlighted the verification of acceptance criteria in various, normal and emergency conditions with large margins. Fatigue analysis has also been conducted. For underclad cracks located in the inner radius of RPV inlet nozzles, a small propagation in the austenitic cladding is demonstrated but would not lead to any cladding perforation within 40 years.

scholarly journals Analysis of O-Ring Seal Failure under Static Conditions and Determination of End-of-Lifetime Criterion

Polymers ◽  
2019 ◽  
Vol 11 (8) ◽  
pp. 1251 ◽  
Author(s):  
Anja Kömmling ◽  
Matthias Jaunich ◽  
Payam Pourmand ◽  
Dietmar Wolff ◽  
Mikael Hedenqvist
Keyword(s):  
Material Properties ◽  
Safety Margin ◽  
Mechanical Analysis ◽  
Butadiene Rubber ◽  
Dimensional Changes ◽  
Seal Failure ◽  
Ring Seal ◽  
Static Conditions

Determining a suitable and reliable end-of-lifetime criterion for O-ring seals is an important issue for long-term seal applications. Therefore, seal failure of ethylene propylene diene rubber (EPDM) and hydrogenated nitrile butadiene rubber (HNBR) O-rings aged in the compressed state at 125 °C and at 150 °C for up to 1.5 years was analyzed and investigated under static conditions, using both non-lubricated and lubricated seals. Changes of the material properties were analyzed with dynamic-mechanical analysis and permeability experiments. Indenter modulus measurements were used to investigate DLO effects. It became clear that O-rings can remain leak-tight under static conditions even when material properties have already degraded considerably, especially when adhesion effects are encountered. As a feasible and reliable end-of-lifetime criterion for O-ring seals under static conditions should include a safety margin for slight dimensional changes, a modified leakage test involving a small and rapid partial decompression of the seal was introduced that enabled determining a more realistic but still conservative end-of-lifetime criterion for an EPDM seal.

Analysis of methodologies for determination of DEM input parameters

2021 ◽  
Author(s):  
Rondinelli M. Lima ◽  
Rodolfo J. Brandao ◽  
Raphael L. Santos ◽  
Claudio R. Duarte ◽  
Marcos A. S. Barrozo
Keyword(s):  
Input Parameters

Experimental Program to Monitor the Irradiation Induced Embrittlement of the French Reactor Pressure Vessel Steels

2004 ◽  
Author(s):  
Guillaume Chas ◽  
Nathalie Rupa ◽  
Josseline Bourgoin ◽  
Astrid Hotellier ◽  
Se´bastien Saillet
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Mechanical Tests ◽  
Nuclear Industry ◽  
Experimental Program ◽  
Surveillance Program ◽  
Base Metals ◽  
Vessel Steel ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

By monitoring the irradiation-induced embrittlement of materials, the Pressure Vessel Surveillance Program (PVSP) contributes to the RPV integrity and lifetime assessments. This program is implemented on each PWR Unit in France; it is mainly based on Charpy tests, which are widely used in the nuclear industry to characterize the mechanical properties of the materials. Moreover, toughness tests are also carried out to check the conservatism of the PVSP methodology. This paper first describes the procedure followed for the Pressure Vessel Surveillance Program. It presents the irradiation capsules: the samples materials (low alloy Mn, Ni, Mo vessel steel including base metals, heat affected zones, welds and a reference material) and the mechanical tests performed. Then it draws up a synthesis of the analysis of about 180 capsules removed from the reactors at fluence levels up to 7.1019 n/cm2 (E > 1 MeV). This database gathers the results of more than 10,000 Charpy tests and 250 toughness tests. The experimental results confirm the conservatism of the Code-based methodology applied to the toughness assessment.

Device Modeling of a-Si:H Alloy Solar Cells: Calibration Procedure for Determination of Model Input Parameters

1998 ◽  
Vol 507 ◽  
Author(s):  
M. Zeman ◽  
R.A.C.M.M. Van Swaaij ◽  
E. Schroten ◽  
L.L.A. Vosteen ◽  
J.W. Metselaar
Keyword(s):  
Solar Cell ◽  
Material Properties ◽  
Calibration Procedure ◽  
Model Parameters ◽  
Model Input ◽  
A Value ◽  
Input Parameters ◽  
Simulated Material ◽  
Good Agreement

ABSTRACTA calibration procedure for determining the model input parameters of standard a-Si:H layers, which comprise a single junction a-Si:H solar cell, is presented. The calibration procedure consists of: i) deposition of the separate layers, ii) measurement of the material properties, iii) fitting the model parameters to match the measured properties, iv) simulation of test devices and comparison with experimental results. The inverse modeling procedure was used to extract values of the most influential model parameters by fitting the simulated material properties to the measured ones. In case of doped layers the extracted values of the characteristic energies of exponentially decaying tail states are much higher than the values reported in literature. Using the extracted values of model parameters a good agreement between the measured and calculated characteristics of a reference solar cell was reached. The presented procedure could not solve directly an important issue concerning a value of the mobility gap in a-Si:H alloys.

scholarly journals Determination of GTN Model Parameters Based on Artificial Neutral Network for a Ductile Failure

2021 ◽  
Vol 2 (1) ◽  
pp. 01-05
Author(s):  
YASSINE CHAHBOUB ◽  
SZAVAI Szabolcs
Keyword(s):  
Ductile Failure ◽  
Compact Tension ◽  
Model Parameters ◽  
Single Edge ◽  
Direct Identification ◽  
Simulation Data ◽  
Short Period ◽  
Artificial Neural Network Ann ◽  
Excellent Tool

The Gurson – Tvergaard – Needleman (GTN) mechanical model is widely used to predict the failure of materials based on laboratory specimens, direct identification of Gurson – Tvergaard – Needleman parameters is not easy and time-consuming, and the most used method to determine them is the combination between the experimental results and those of the finite elements, the process consists of repeating the simulations several times until the simulation data matches the experimental data obtained at the specimen level.This article aims to find GTN parameters for the Compact Tension (CT) and Single Edge Tensile Test (SENT) specimen based on the Notch Specimen (NT) using the Artificial Neural Network (ANN) approach. . This work presents how the ANN could help us determine the parameters of GTN in a very short period of time. The results obtained show that ANN is an excellent tool for determining GTN parameters.

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