Qualitative Experimental Study About the Safety Injection in the Cold Leg of a Reactor Pressure Vessel

2003 ◽  
Author(s):  
Th. Bichet ◽  
A. Martin ◽  
F. Beaud
Keyword(s):  
Experimental Study ◽  
Pressure Vessel ◽  
Single Phase ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Two Phase ◽  
Hydraulic Behavior ◽  
Plant Life ◽  
Physical Phenomena ◽  
Reactor Pressure

This paper presents the experimental results of a qualitative study concerning the physical phenomena present in a cold leg and the downcomer of a reactor pressure vessel during a safety injection scenario. This project contributes to the plant life time project. Single-phase and two-phase scenarios have been studied according to the thermal-hydraulic behavior obtained by system codes. This paper shows different physical phenomena visualisations concerning the behavior of the fluid flow at different location in downcomer and in cold leg and physical phenomena in a uncovered cold leg.

CFD Tools for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Thermalhydraulic Benchmark on the Validation Test Case UPTF

2010 ◽  
Author(s):  
A. Martin ◽  
F. Huvelin ◽  
G. Balard ◽  
S. Bellet ◽  
B. Durand ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Single Phase ◽  
Physical Phenomenon ◽  
Reactor Pressure Vessel ◽  
Test Facility ◽  
Test Case ◽  
Phase Flow ◽  
Two Phase ◽  
Reactor Pressure

The Reactor Pressure Vessel (RPV) assessment and lifetime evaluation of the nuclear plants leads EDF and AREVA to use CFD tools for the demonstration of the Reactor Pressure Vessel (RPV) integrity. French PWR plants have been studied for a long time and so a large number of computations have been performed in three steps: System — CFD — Mechanical Analysis. This paper focuses on the CFD aspects and especially on the two different CFD tools imposed by the physical phenomena of the transient scenarii. Indeed, the results of the system code analysis (CATHARE computations) of the PTS transient induce two kinds of scenarii: single phase and two-phase flows in the cold leg. According to the approach considered, it was interesting to carry out a global benchmark with the different thermalhydraulic tools. In this frame, Code_Saturne and Star CD CFD tools used for single phase flow scenario configuration and NEPTUNE_CFD used for two phase flow configuration were confronted with UPTF (Upper Plenum Test Facility) experiment results in a single phase scenario. The choice to perform this benchmark on UPTF test case was retained because this integral test was very representative of the main physical phenomenon that is well suited for the validation of the CFD tools used to study the Pressurised Thermal Shock for the RPV integrity.

Stratified Flow and Free Surface of an Uncovered Reactor Pressure Vessel Cold Leg: Water-Air Experimental Results

2005 ◽  
Author(s):  
Thierry Bichet ◽  
Alain Martin ◽  
Fre´de´ric Beaud
Keyword(s):  
Free Surface ◽  
Pressure Vessel ◽  
Two Phase Flow ◽  
Salt Water ◽  
Stratified Flow ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Phase Flow ◽  
Two Phase ◽  
Reactor Pressure

Within the Reactor Pressure Vessel (RPV) plant life time project, Electricite´ de France (EDF) must carry out studies showing the RPV integrity during severe loading conditions as the Small Break Loss Of Coolant Accidents (SBLOCA) scenarios. A recent thermal-hydraulic SBCLOCA scenario study, concerning a certain RPV geometry, showed two-phase-flow conditions appears in the cold leg. The single phase flow CFD tools, usually used in these cases, are unsuitable to carry out such studies. That is the reason why EDF is developing a new CFD code (called Neptune_Code) implementing two-phase flow models. In order to improve the physical phenomena knowledge and the existent experimental data for Neptune_Code development, EDF carried out an experimental study concerning the stratified flow (salt water) and free surface (water air) behavior during a representative ECC scenario with partial uncovered cold leg. This paper focuses mainly on the experimental method and the measurements obtained in three tests showing the effects of the uncovered cold leg level on the stratified flow and the free surface behavior.

EPR: Tests Performed to Confirm the Mechanical and Hydraulic Design of the Vessel Internals

2008 ◽  
Author(s):  
Philippe Dolleans ◽  
Charlotte de Monplanet ◽  
Jean-Philippe Fontes
Keyword(s):  
Pressure Vessel ◽  
Operating Experience ◽  
Forthcoming Paper ◽  
Reactor Pressure Vessel ◽  
Plant Design ◽  
Control Rod ◽  
Engineering Teams ◽  
Hydraulic Behavior ◽  
Under Construction ◽  
Reactor Pressure

The EPR is an Evolutionary high-Power Reactor which is based on the best French and German experience of the past twenty years in plant design construction and operation. In the present detailed engineering phase of the plant under construction in Finland (Okiluoto 3) and in France (Flamanville 3), some actions were conducted in order to improve the knowledge of the hydraulic behavior of the innovative Reactor Pressure Vessel internals (RPV). The RPV internals are mainly derived from former French N4 or German Konvoi with some evolutions to take into account the operating experience. Design and validation of the internals were performed within AREVA’s engineering teams, which develop state of the art methods in the field of thermohydraulic testing. The experimental validation program was closely followed by EDF. Moreover, an EDF R&D project, whose results are not addressed here, was held to consolidate the RPV internals conception. The aim of the paper is to present the hydraulic tests performed on mock-ups to characterize the hydraulic behavior of the innovative EPR Reactor Pressure Vessel internals, and to introduce the role of these tests in the global conception process of the EPR RPV internals (CFD code qualification, design validation, database...). The qualification of the CFD computer codes will be described in a forthcoming paper. Three different mock-ups are presented to illustrate these tests: • JULIETTE for the reactor pressure vessel lower internals, • ROMEO for the reactor pressure vessel upper internals, • MAGALY for the design of the skeleton-type control rod guide assembly.

CFD-Tools for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Thermal-Hydraulic Methods and Main Industrial Results on the French 900 MWe PWR RPV

2009 ◽  
Author(s):  
A. Martin ◽  
F. Lestang ◽  
S. Bellet ◽  
D. Guichard ◽  
C. Vit ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Fluid Temperature ◽  
Time Duration ◽  
Two Phase ◽  
Safety Margins ◽  
Reactor Pressure ◽  
Fast Fracture

For the Reactor Pressure Vessel (RPV) assessment and lifetime evaluation of the nuclear plants, French Utilities apply a series of calculations including thermal-hydraulic, thermo mechanical and fracture mechanics studies in order to study the Pressurized Thermal Shock (PTS) in the downcomer caused by the safety injection. Within the frame of the plant lifetime project, integrity assessments of the French 900 MWe (3-loops) series RPV have been performed. A gain for safety margins to fast fracture of the RPV can be found with a 3D modeling of thermal-hydraulics loads. From a physical phenomena point of view, the results of the system code analysis (CATHARE computation) of the PTS transient induce two kinds of scenarios: single phase and two-phase flows in the cold leg. In the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. For that purpose, a program has been set up to extend the capabilities of the NEPTUNE_CFD two-phase solver which is the tool able to solve two-phase flow configuration. At the same time, a simplified approach has shown that for this kind of scenario where the cold leg is weakly uncovered, a free surface calculation (without phase change) was sufficient to respect the necessary criteria of safety. Considering the time duration of 3D computation and the large number of cases, EDF and AREVA-NP decided to share the effort. The two teams use the NEPTUNE_CFD code (coupled with the thermal solid SYRTHES code) for thermal-hydraulic computations. The thermo mechanical code used is CALORI. According to this approach and to reduce the CPU time, two computations have been performed for 2″ and 3″ Small Break Loss Of Coolant Accident (SBLOCA) on a one-third RPV model. Computations on a complete RPV model have been performed to demonstrate the relevance of the one-third RPV model. The studies have been performed by two independent teams from EDF and AREVA-NP. The investigated configuration corresponds to the injection of cold water in the RPV during a penalizing representation of a primary break transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of fluid temperature in the cold legs and in the downcomer. The obtained numerical description of the transient is used as boundary conditions for a full mechanical computation of the stresses. The results show that such a complete thermal-hydraulic and mechanical 3-dimensional analysis improves the evaluation of the consequences of the loading on the stress fields and eventually the margins to fast fracture of the RPV. The good agreement observed between a one-third RPV model and a complete RPV model results confirms the validity of the approach.

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

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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