scholarly journals The Cause of Intergranular (IG) Fracture by Thermal Embrittlement in SA508 of Reactor Pressure Vessel (RPV) Steel

2021 ◽  
Vol 59 (9) ◽  
pp. 589-601
Author(s):  
Sung Soo Kim ◽  
Jung Jong Yeob ◽  
Young Suk Kim
Keyword(s):  
Thermal Treatment ◽  
Pressure Vessel ◽  
Boundary Segregation ◽  
Entropy Change ◽  
Reactor Pressure Vessel ◽  
Atomic Diffusion ◽  
New Paradigm ◽  
Lattice Contraction ◽  
Rpv Steel ◽  
Reactor Pressure

Intergranular(IG) fracture due to thermal treatment has been reported in a reactor pressure vessel(RPV) steel of Russian light water reactor in last decade. This is attributed to grain boundary segregation of phosphorus (P) or precipitation of carbide, etc.. This is a finding a difference in microstructure before and after IG cracking; this cannot explain the cause of the IG embrittlement. This old paradigm follows only correlation. Recently, a mechanism in which IG embrittlement occurs due to a decrease in entropy of a material has been reported at a temperature where atomic diffusion is possible. It is anticipated that new paradigm can explain the IG embrittlement of RPV based on a causal relationship. Thus, the thermal treatment at 350-420 oC was applied to RPV steel of SA508 and IG cracking was confirmed. DSC analysis was applied to confirm whether a decrease in entropy due to a short range ordering reaction occurs in SA508. It was possible to quantify the entropy change(⊿S= Q/T) through DSC measurement. A lattice changes due to thermal treatment were confirmed using XRD analysis in aged specimens. The results showed that lattice contraction by aging causes a reduction of fracture toughness. The internal stress formed inside the material due to entropy reduction can be calculated by multiplying the exothermic energy per unit mass by the density. This relationship is expressed by a equation of stress(σ) = exothermic heat(⊿Q) x density(ρ).

Effects of Neutron Irradiation on the Mechanical and Electromagnetic Properties of Reactor Pressure Vessel Steels

2017 ◽  
Author(s):  
Li Chengliang ◽  
Shu Guogang ◽  
Chen Jun ◽  
Liu Yi ◽  
Liu Wei ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Plants ◽  
Hysteresis Loops ◽  
Reactor Pressure Vessel ◽  
Electromagnetic Properties ◽  
Irradiation Damage ◽  
Indentation Hardness ◽  
Rpv Steel ◽  
Reactor Pressure

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.

Nondestructive SANS Study of Reactor Pressure Vessel Steel after Irradiation

2006 ◽  
Vol 321-323 ◽  
pp. 1667-1670
Author(s):  
Young Soo Han ◽  
Eun Joo Shin ◽  
Baek Seok Seong ◽  
Chang Hee Lee ◽  
Duck Gun Park
Keyword(s):  
Neutron Scattering ◽  
Small Angle ◽  
Pressure Vessel ◽  
Small Angle Neutron Scattering ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Rpv Steel ◽  
Induced Defects ◽  
Irradiation Induced Defects ◽  
Reactor Pressure

The irradiation induced defects of irradiated reactor pressure vessel(RPV) steel were investigated by a small angle neutron scattering. The degradation of the mechanical properties of RPV steels during an irradiation in a nuclear power plant is closely related to the irradiation induced defects. The size of these defects is known to be a few nanometers, and the small angle neutron scattering technique is regarded as the best non destructive technique to characterize the nano sized inhomogeneities in bulk samples. The RPV steel was irradiated in the HANARO reactor in KAERI. The small angle neutron scattering experiments were performed at SANS instrument in the HANARO reactor. Both unirradiated and irradiated RPV steels were measured and the SANS data of both steels were compared. The nano sized irradiation induced defects were quantitatively analyzed by SANS. The type of defects was also analyzed based on the SANS results, and the effect of the chemical composition of the RPV steel on the irradiation induced defects was discussed.

State of Advancement of the International REVE Project: Computational Modelling of Irradiation-Induced Hardening in Reactor Pressure Vessel Steels and Relevant Experimental Validation Programme

2002 ◽  
Author(s):  
Lorenzo Malerba ◽  
Eric van Walle ◽  
Christophe Domain ◽  
Stephanie Jumel ◽  
Jean-Claude Van Duysen
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Methodological Approach ◽  
Computational Modelling ◽  
Ternary Systems ◽  
Reactor Pressure Vessel ◽  
Dislocation Loops ◽  
Simulation Tools ◽  
Rpv Steel ◽  
Reactor Pressure

The REVE (REactor for Virtual Experiments) project is an international joint effort aimed at developing multiscale modelling computational toolboxes capable of simulating the behaviour of materials under irradiation at different time and length scales. Well grounded numerical techniques such as molecular dynamics (MD) and Monte Carlo (MC) algorithms, as well as rate equation (RE) and dislocation-defect interaction theory, form the basis on which the project is built. The goal is to put together a suite of integrated codes capable of deducing the changes in macroscopic properties starting from a detailed simulation of the microstructural changes produced by irradiation in materials. To achieve this objective, several European laboratories are closely collaborating, while exchanging data with American and Japanese laboratories currently pursuing similar approaches. The material chosen for the first phase of this project is reactor pressure vessel (RPV) steel, the target macrosocopic magnitude to be predicted being the yield strenght increase (Δσy) due, essentially, to irradiation-enhanced formation of intragranular solute atom precipitates or clouds, as well as irradiation induced defects in the matrix, such as point defect clusters and dislocation loops. A description of the methodological approach used in the project and its current state is given in the paper. The development of the simulation tools requires a continuous feedback from ad hoc experimental data. In the framework of the REVE project SCK·CEN has therefore performed a neutron irradiation campaign of model alloys of growing complexity (from pure Fe to binary and ternary systems and a real RPV steel) in the Belgian test reactor BR2 and is currently carrying on the subsequent materials characterisation using its hot cell facilities. The paper gives the details of this experimental programme — probably the first large-scale one devoted to the validation of numerical simulation tools — and presents and discusses the first available results, with a view to their use as feedback for the improvement of the computational modelling.

Fracture Toughness Evaluation of Neutron-Irradiated Reactor Pressure Vessel Steel Using Miniature-C(T) Specimens

2017 ◽  
Author(s):  
Yoosung Ha ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Yutaka Nishiyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Rpv Steel ◽  
Toughness Evaluation ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure

The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. By carefully selecting the test temperature, valid fracture toughness and reference temperature (To) were determined successfully with a relatively small number of specimens. The value of To determined using irradiated Mini-C(T) specimens was in good agreement with that determined using irradiated pre-cracked Charpy-type (PCCv) specimens. In addition, the scatter of the 1T-equivalent fracture toughness values obtained using the irradiated Mini-C(T) specimens was not significantly different from that obtained using the irradiated PCCv and other larger unirradiated specimens. The To values determined using Mini-C(T) specimens agree very well with the correlation between the Charpy 41J transition temperature and the To of commercially manufactured RPV steels reported in the past.

Constraint Effect on Fracture Mechanics Evaluation for an Under-Clad Crack in a Reactor Pressure Vessel Steel

2020 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Satoshi Hanawa
Keyword(s):  
Fracture Toughness ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Constraint Effect ◽  
Rpv Steel ◽  
Low Constraint ◽  
Reactor Pressure

Abstract For structural integrity assessment of the reactor pressure vessel (RPV) in JEAC 4206-2016, it is required that the fracture toughness (KJc) be higher than the stress intensity factor at the crack tip of a postulated under-clad crack (UCC) near the inner surface of RPV steel under the pressurized thermal shock event. Previous analytical studies showed a low constraint effect at the crack tip of an UCC, compared with that of a normal surface crack. Such a low constraint effect may increase the apparent KJc. In this study, we performed three-point bending (3PB) fracture toughness tests and finite element analysis (FEA) for RPV steel containing an UCC or a surface crack to quantitatively investigate the effect of cladding on the KJc. The FEAs considering the anisotropic property of the cladding successfully reproduced the load vs. load-line displacement curves obtained from the tests. We found that the apparent KJc for the UCC was considerably higher than that for the surface crack. FEA also showed that the constraint effect for the 3PB test specimen with the UCC was lower than that for the specimen with the surface crack owing to the cladding. Thus, a low constraint effect from an UCC may increase the apparent KJc.

Toughness and Embrittlement of RPV Steels Using Ultrasonic Measurements

2004 ◽  
Author(s):  
Allen L. Hiser
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Ultrasonic Attenuation ◽  
Reactor Pressure Vessel ◽  
Irradiation Embrittlement ◽  
Nondestructive Determination ◽  
Rpv Steel ◽  
Suitable Parameter ◽  
Reactor Pressure

Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper summarizes the findings from an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. The materials examined in this study include one heat of RPV steel that was heat treated to induce changes in its fracture toughness, several heats of RPV steel irradiated to assess neutron embrittlement changes in fracture toughness, and a matrix of RPV steels (in the unirradiated condition) with a range of as-fabricated fracture toughness levels.

Reactor Pressure Vessel: EDF R&D Program to Support Lifetime Management

2011 ◽  
Author(s):  
Jean-Philippe Fontes ◽  
Christelle Raynaud ◽  
Alain Martin ◽  
Aurore Parrot ◽  
Anna Dahl ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Crack Arrest ◽  
Reactor Pressure Vessel ◽  
Vessel Steel ◽  
High Fluences ◽  
Rpv Steel ◽  
Safety Studies ◽  
Key Issues ◽  
Physical Phenomena ◽  
Reactor Pressure

The Reactor Pressure Vessel (RPV) is with the concrete containment one of the two components of a NPP whose replacement is not considered as reasonably feasible. The RPV lifetime has thus an important impact on the lifetime of the whole NPP. One of the key issues concerning RPV lifetime is the radiation effect on the RPV steel in the core zones. The vessel steel becomes indeed more brittle in the RPV core region where radiation is high. Margins have been included at design and manufacturing stages taking into account the material’s embrittlement. Moreover, operating measures have been taken to manage ageing of RPV in order to extend lifetime. The challenge is to preserve high margins and to provide the safety studies showing these margins. A large R&D program has been developed to support lifetime extension. The objective of the program is to develop tools and provide input data for the demonstration of the safe operation of the reactor pressure vessel significantly over a 40-year lifetime. The aim of the paper is to present an overview of the R&D program to support lifetime management on the fields of materials, mechanics and thermalhydraulics. Experiments are indeed performed on irradiated material in order to improve the knowledge on embrittlement for high fluences and to be able to determine embrittlement correlations for materials representative of French RPV. Actions are also planned to improve evaluation of the RPV mechanical behaviour and to describe physical phenomena such as crack arrest or warm pre-stressing effect. Last, studies are realized to improve the thermal loadings evaluations under hypothetical accidental scenarios. These studies are supported by thermalhydraulic numerical simulations whose validation is obtained by comparison to experimental results from experimental hydraulic loops representative of French RPV.

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