An application of Charpy V testing: The pressure vessel surveillance program of nuclear pressurised water reactor in operation

2002 ◽  
pp. 365-375
Author(s):  
Nathalie Rupa ◽  
Astrid Baché ◽  
Josseline Bourgoin ◽  
Denis Buisine
Keyword(s):  
Pressure Vessel ◽  
Surveillance Program ◽  
Water Reactor

The BWRVIP Integrated Surveillance Program

2004 ◽  
Author(s):  
Robert G. Carter ◽  
Timothy J. Griesbach ◽  
Timothy C. Hardin
Keyword(s):  
Research Program ◽  
Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Surveillance Program ◽  
Water Reactor ◽  
Weld Material ◽  
Surveillance Programs ◽  
Reactor Pressure ◽  
The U.S

Boiling Water Reactor (BWR) plants in the U.S. are designed with radiation surveillance programs. However, the surveillance materials in some plants do not necessarily represent the limiting plate and/or weld material of the reactor pressure vessel (RPV). Also, some plants do not have baseline data for the surveillance materials, which is needed to measure irradiation shift. In 1998 the BWR Vessel and Internals Project (BWRVIP) conceived the BWR Integrated Surveillance Program (ISP) to address these concerns. The ISP surveyed all BWR vessel limiting materials and all available BWR surveillance materials (including materials from a 1990s supplementary research program called the Supplemental Surveillance Program, or SSP). For each vessel limiting weld and limiting plate, a best representative surveillance material was assigned, based on heat number, similar chemistries, common fabricator, and the availability of unirradiated data. Many of the selected surveillance materials are good representatives for the limiting materials of multiple plants, so fewer capsules are required to be tested, reducing the overall cost of surveillance while also improving BWR fleet compliance with 10CFR50 Appendix H.

Boiling Water Reactor (BWR) Pressure Vessel Surveillance Program

2008 ◽  
pp. 106-106-12
Author(s):  
KS Sivaramakrishnan ◽  
S Chatterjee ◽  
S Anantharaman ◽  
KS Balakrishnan ◽  
UK Viswanathan ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Surveillance Program ◽  
Water Reactor

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Experience on ultrasonic inspection of control rod nozzles of pressurised water reactor pressure vessel

2007 ◽  
Vol 2 (2) ◽  
pp. 129-132
Author(s):  
P. Kauppinen ◽  
H. Jeskanen ◽  
R. Paussu ◽  
B. Elsing
Keyword(s):  
Pressure Vessel ◽  
Ultrasonic Inspection ◽  
Reactor Pressure Vessel ◽  
Control Rod ◽  
Water Reactor ◽  
Reactor Pressure

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.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

Sign in / Sign up

Export Citation Format

Share Document