In Electricite´ de France (EDF), it’s used to taking as a conservative constant value for the temperature of Refueling Water Storage Tank (RWST) in the deterministic and probabilistic analyses for Reactor Pressure Vessel (RPV) life management. The water contained in this storage tank supplies Security Injection during accidental conditions, such as loss-of-coolant accident (LOCA), so the temperature of this water is a very important input parameter for fracture mechanics analyses. In the continuation of [1], the aim of our study is to evaluate the variability of this temperature. Since 1999, EDF has been collecting the RWST temperature for several sites and several nuclear plant units. Using statistical analyses, this study aims at first to identify the most important RPV exploitation life events that influence this temperature and, at secondly, to obtain statistical density distributions in order to describe its variability in the most realistic way possible.

This paper explains the Research and Development program started at E.D.F about the cooling phenomena of a PWR vessel after a Pressurized Thermal Shock. The numerical results are obtained with the thermalhydraulic code Code_Saturne coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. The geometry used represents a four loop PWR plant. In this calculation, the simulated geometry takes into account as much as possible the exact geometry of the lower plenum such as its columns and plates instrumentation. The configuration investigated is related to the injection of cold water in the vessel during a penalizing operating transient and its impact on the solid part formed by cladding and base metal. Numerical results are given in terms of temperature field in the cold legs and in the down comer. The obtained numerical description of the transient (internal pressure and temperature field within the vessel) is used as boundary conditions for a full mechanical computation of the stresses. This thermal–mechanical transient is obtained by F.E. simulation using the F.E. code Code_Aster on a 3D mesh of the vessel, covering the two core–shells and their circumferential welds, as well as the internal cladding. Based on an analytical method specially established for underclad flaws, the corrected stress intensity factor Kβ during the transient is evaluated for an hypothetical flaw, by extracting the stresses along a radial segment. The severity of the flaw with respect to the transient is quantified by the minimum of the ratio KIc/Kβ, where KIc refers to the base metal fracture toughness for brittle initiation. The evolution of the severity with the position of the hypothetical flaw is studied and compared with the results given by a classical uni–dimensional method. The results show that such a complete thermal–hydraulic and mechanical 3–dimensional analysis allows to reduce considerably the severity of the flaws, thus improving the integrity of the RPV.

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.

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behaviour of cracks under PTS loading conditions due to the emergency cooling during PTS transient like SBLOCA. This paper explains the Research and Development program started at Electricite´ De France about the cooling phenomena of a PWR vessel after a Pressurised Thermal Shock. The numerical results are obtained with the E.D.F ThermalHydraulic code (Code_Saturne) coupled with the thermal-solid code SYRTHES to take into account the conjugate heat transfer on the cooling of the vessel. We first explain the global methodology with a progress report on the state of the art of the tools available to simulate the different scenari displayed within the frame of the plant life project in order to reassess the integrity of the RPV, taking into account the evolution of some input data, such as the new value of end of life (EOL) fluence, the feedback results of surveillance program and the evolution of the functional requirements. The main results are presented and are related to the evaluation of the RPV integrity during a Small Break Loss Of Coolant Accident transient for 900 and 1300 MWe nuclear plant. On the whole, the main purpose of the numerical CFD studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margins. In a second time, a new analysis is performed to assess an accurate temperature distribution in the RPV. Indeed, from a physical phenomena point of view, the EDF thermalhydraulic tool Code_Saturne is now qualified in order to assess single phase transient but in the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the emergency core cooling injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3D codes. In 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. In a same time, A simplified approach has showed that for a type of transient weakly uncovered, a free surface calculation was sufficient to respect the necessary criteria of safety. A Qualification study was carried out on the Hybiscus experimental E.D.F facility, representing a cold leg with ECC injection and a third down comer. Temperature profiles have been compared and are presented and analysed here, showing encouraging results.

Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local thermal-hydraulic analysis of a small break loss of coolant accident transient, this paper presents mainly a parametric study that helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel, and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation, which will subsequently assess the associated RPV safety margin factors.