CFD Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2011 ◽  
Vol 133 (3) ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

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.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2006 ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

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 behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, 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, the paper presents mainly a parametric study which 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 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 margin factors.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling

2005 ◽  
Author(s):  
A. Martin ◽  
D. Monfort ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

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 behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, 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, the paper presents mainly a parametric study whose purpose is to understand the main phenomena which 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. On the whole, the main purpose of the numerical thermal-hydraulic 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 margin factors.

Re-Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2015 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hu Hui ◽  
Hui Li ◽  
Fuzhen Xuan
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Master Curve Method ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RTPTS values were determined by the tangent approach. In the case that the RTNDT at or beyond the RPV design life may exceed the RTPTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RTPTS values are increased to varied extents for the different loading transients.

Thermal Hydraulic Analysis of Pressurized Thermal Shock for Loviisa NPP Using Apros Simulation Code

2018 ◽  
Author(s):  
Gintaras Žemulis ◽  
Pasi Junninen ◽  
Petri Kytömäki
Keyword(s):  
Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Large Set ◽  
Energy Company ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure ◽  
Emergency Operating

Loviisa Nuclear Power Plant consists of two VVER-440 type pressurized water reactor units and it is owned by a Finnish energy company Fortum Power and Heat Oy. Only some years after Loviisa Unit 1 start-up, during early 1980’s, some reactor pressure vessel embrittlement by fast neutron irradiation was recognized. Embrittlement of the reactor pressure vessel materials increases the risk of a fast fracture under emergency core cooling conditions. This led to an extensive study of pressurized thermal shock (PTS) scenarios, which included experimental research, large set of probabilistic and deterministic safety analyses, code development and plant modifications to both lower overcooling transient probability and to mitigate it’s consequences. Current operating licenses for Loviisa 1 and 2 Units reactor pressure vessels were applied by Fortum in 2012 to cover operation up to the years 2027 and 2030, respectively. The operating licenses were granted as applied by the Finnish Radiation and Nuclear Safety Authority (STUK) according to application. However a know-how of PTS scenarios shall be maintained. This becomes even more important in a light of any big power plant automation modernization projects targeting also safety systems. Currently Fortum is involved in an automation renewal project also known under its ELSA-project name. The project is scheduled to be ready in 2018. The project targets at replacing parts of old analogue automation with new digital systems and also extending their diversity by introducing new safety functions for emergency management scenarios. As a result, changes to some emergency operating procedures are inevitable. Due to the changes to emergency operating procedures PTS-scenarios for Loviisa NPP have been also updated. In 2016 Fortum launched PTS-project in connection with ELSA-project. Previous PTS-scenario related thermal hydraulic analyses were based on the original 80’s decade analyses, which were subsequently revised many times following other modernization projects that took place at Loviisa NPP. In this new round of PTS-analyses the most important cases were recalculated using Apros (Advanced Process Simulator). ELSA-project introduced changes are not only emergency procedures specific but also involve safety system related process and automation logic modifications. The changes affected therefore boundary conditions needed in the analyses. In total probabilistic safety analyses included over 160 different PTS-sequences. The most limiting sequences were selected for deterministic analyses. PTS-analyses were performed in co-operation with Platom Oy and plant owner Fortum Power and Heat Oy. The main results of PTS-analyses are presented in the paper.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

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