This paper describes the analytical work done by modelling and evaluating a thermal shock in a WWER-440 reactor pressure vessel due to an emergency case. An axial oriented semielliptical underclad/surface crack is assumed to be located in the core weld line. Threedimensional finite element models are used to compute the global transient temperature and stress-strain fields. By using a three-dimensional submodel, which includes the crack, the local crack stress-strain field is obtained. With a subsequent postprocessing using the j-integral technique the stress intensity factors KI along the crack front are obtained. The results for the underclad and surface crack are provided and compared, together with a critical discussion of the VERLIFE c...
The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) wi...
One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressuris...
In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during p...
This paper describes the analytical work done by modelling and evaluating a thermal shock in a WWER-...
Comparative analyses have been performed with 3-D finite element (FE) models for a six loop cladded ...
The considered research activity deals with the application of a chain of numerical codes, in order ...
A joint pressure vessel integrity research programme involving three partners is being carried out d...
In order to simulate a nuclear emergency cooling situation, longterm cooling tests (pressurized ther...
In order to simulate a nuclear emergency cooling situation, longterm cooling tests (pressurized ther...
ABSTRACT The cladding of an RPV (Reactor Pressure Vessel) at the inner surface may be subjected to a...
In the paper, evaluation of 9 large scale experiments performed on uncladded cracked beams made of R...
Since about a decade an important issue appeared in nuclear technology and in nuclear reactor safety...
Since about a decade an important issue appeared in nuclear technology and in nuclear reactor safety...
The Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pr...
Three-dimensional J-integral and two-dimensional Local Approach finite element studies are described...
The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) wi...
One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressuris...
In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during p...
This paper describes the analytical work done by modelling and evaluating a thermal shock in a WWER-...
Comparative analyses have been performed with 3-D finite element (FE) models for a six loop cladded ...
The considered research activity deals with the application of a chain of numerical codes, in order ...
A joint pressure vessel integrity research programme involving three partners is being carried out d...
In order to simulate a nuclear emergency cooling situation, longterm cooling tests (pressurized ther...
In order to simulate a nuclear emergency cooling situation, longterm cooling tests (pressurized ther...
ABSTRACT The cladding of an RPV (Reactor Pressure Vessel) at the inner surface may be subjected to a...
In the paper, evaluation of 9 large scale experiments performed on uncladded cracked beams made of R...
Since about a decade an important issue appeared in nuclear technology and in nuclear reactor safety...
Since about a decade an important issue appeared in nuclear technology and in nuclear reactor safety...
The Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pr...
Three-dimensional J-integral and two-dimensional Local Approach finite element studies are described...
The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) wi...
One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressuris...
In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during p...