Probabilistic Pressurized Thermal Shock Analysis for a Reactor Pressure Vessel Considering Plume Cooling Effect

2016 ◽  
Vol 138 (4) ◽  
Author(s):  
Guian Qian ◽  
V. F. González-Albuixech ◽  
Markus Niffenegger ◽  
Medhat Sharabi
Keyword(s):  
Crack Initiation ◽  
Pressure Vessel ◽  
Cooling Water ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Reactor Pressure ◽  
Thermal Shocks

The inner surface of a reactor pressure vessel (RPV) is assumed to be subjected to pressurized thermal shocks (PTSs) caused by the injection of emergency cooling water. The downstream is not homogeneous but typically in a plume shape coming from the inlet nozzles. In this paper, both deterministic and probabilistic methods are used to assess the integrity of a model RPV subjected to PTS. The favor code is used to calculate the probabilities for crack initiation and failure of the RPV considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs. The study shows that peak KI of the cracks inside the plume increases about 33% compared with that outside. The conditional probability inside the plume is more than eight orders of magnitude higher than outside the plume. In order to be conservative, it is necessary to consider the plume effect in the integrity assessment.

Probabilistic Pressurized Thermal Shocks Analysis for a Reactor Pressure Vessel by Considering Plume Effect

2015 ◽  
Author(s):  
Guian Qian ◽  
V. F. González-Albuixech ◽  
Markus Niffenegger ◽  
Medhat Sharabi
Keyword(s):  
Crack Initiation ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Cooling Water ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Integrity Assessment ◽  
Inner Surface ◽  
Reactor Pressure ◽  
Thermal Shocks

The inner surface of a reactor pressure vessel (RPV) is assumed to be subjected to pressurized thermal shocks (PTSs) caused by the downstream of emergency cooling water. The downstream is not homogeneous but typically in a plume shape coming from the inlet nozzles. In this paper, both deterministic and probabilistic methods are used to assess the integrity of a model RPV subjected to PTS. The FAVOR code is used to calculate the probabilities for crack initiation and failure of the RPV considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs. The study shows that peak KI of the cracks inside the plume increases about 33% compared with that outside. The conditional probability inside the plume is more than eight orders of magnitude higher than outside the plume. In order to be conservative, it is necessary to consider the plume effect in the integrity assessment.

Weld Material Investigations of a WWER-440 Reactor Pressure Vessel: Results From the First Trepan Taken From the Former Greifswald NPP

2009 ◽  
Vol 131 (2) ◽  
Author(s):  
Udo Rindelhardt ◽  
Hans-Werner Viehrig ◽  
Joerg Konheiser ◽  
Jan Schuhknecht
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Monte Carlo Code ◽  
Data Set ◽  
Integrity Assessment ◽  
Inner Surface ◽  
Reactor Pressure

Between 1973 and 1990 four units of the Russian nuclear power plants type WWER-440/230 were operated in Greifswald (former East Germany). Material probes from the pressure vessels were gained in the frame of the ongoing decommissioning procedure. The investigations of this material started with material from the circumferential core weld of unit 1. First, this paper presents results of the reactor pressure vessel (RPV) fluence calculations depending on different loading schemes and on the axial weld position based on the Monte Carlo code TRAMO. The results show that the use of the dummy assemblies reduces the flux by a factor of 2–5 depending on the azimuthal position. The circumferential core weld (SN0.1.4) received a fluence of 2.4×1019 neutrons/cm2 at the inner surface; it decreases to 0.8×1019 neutrons/cm2 at the outer surface. The material investigations were done using a trepan from the circumferential core weld. The reference temperature T0 was calculated with the measured fracture toughness values, KJc, at brittle failure of the specimen. The KJc values show a remarkable scatter. The highest T0 was about 50°C at a distance of 22 mm from the inner surface of the weld. The Charpy transition temperature TT41J estimated with results of subsized specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The VERLIFE procedure prepared for the integrity assessment of WWER RPV was applied on the measured results. The VERLIFE lower bound curve indexed with the Structural Integrity Assessment Procedures for European Industry (SINTAP) reference temperature, RTT0SINTAP, envelops the KJc values. Therefore for a conservative integrity assessment the fracture toughness curve indexed with a RT representing the brittle fraction of a data set of measured KJc values has to be applied.

scholarly journals Probabilistic analysis of PWR Reactor Pressure Vessel under Pressurized Thermal Shock

2021 ◽  
Vol 8 (1) ◽  
pp. 1-9
Author(s):  
Kuen Ting ◽  
Anh Tuan Nguyen ◽  
Kuen Tsann Chen ◽  
Li Hwa Wang ◽  
Yuan Chih Li ◽  
...  
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Term Operation ◽  
Oak Ridge ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The beltline region is the most important part of the reactor pressure vessel, become embrittlement due to neutron irradiation at high temperature after long-term operation. Pressurized thermal shock is one of the potential threats to the integrity of beltline region also the reactor pressure vessel structural integrity. Hence, to maintain the integrity of RPV, this paper describes the benchmark study for deterministic and probabilistic fracture mechanics analyzing the beltline region under PTS by using FAVOR code developed by Oak Ridge National Laboratory. The Monte Carlo method was employed in FAVOR code to calculate the conditional probability of crack initiation. Three problems from Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel (PROSIR) round-robin analysis were selected to analyze, the present results showed a good agreement with the Korean participants’ results on the conditional probability of crack initiation.

Probabilistic Pressurized Thermal Shocks Analysis for a Reactor Pressure Vessel

2014 ◽  
Author(s):  
Guian Qian ◽  
Markus Niffenegger
Keyword(s):  
Aspect Ratio ◽  
Pressure Vessel ◽  
Crack Depth ◽  
Crack Density ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Critical Crack ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

Both deterministic and probabilistic methods are used to analyze the integrity of a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code is applied to calculate the probabilities for crack initiation and failure of the RPV subjected to two transients, by considering crack distributions based on cracks observed in the Shoreham and pressure vessel research user facility (PVRUF) RPVs. The crack parameters, i.e. crack density, depth, aspect ratio, orientation and location are assumed as random variables following different distributions. KI of the cracks with the same depth increases with its aspect ratio. Both KI and KIC at the crack tip increase with crack depth, which is the reason why a deeper crack does not necessarily lead to a higher failure probability. The underclad crack is the most critical crack and the deeper crack is the least critical one in this study. Considering uncertainties of the transients results in higher failure probabilities.

Warm Pre-Stressing Tests for WWER 440 Reactor Pressure Vessel Material

2009 ◽  
Author(s):  
Dana Lauerova ◽  
Vladislav Pistora ◽  
Milan Brumovsky ◽  
Milos Kytka
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Point Of View ◽  
Evaluation Procedure ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Model Predictions ◽  
Vessel Material ◽  
Almost All ◽  
Reactor Pressure

During years 2006 – 2008, warm pre-stressing tests on small (Charpy size) and 1T CT specimens were performed at NRI Rez. The specimens were made from WWER 440 reactor pressure vessel material in as-received, thermally treated (artificially aged) and irradiated conditions, the last two conditions simulating the end of life state of the RPV. In this paper, only results of tests performed for this material in as-received and irradiated conditions are presented. Evaluation of WPS tests was performed with using Chell and Wallin predictive models. The attention was paid to 5% probability level fracture predictions, since this level of probability is important for WPS application in pressurized thermal shock evaluation performed within the RPV integrity assessment. From point of view of this 5% probability fracture prediction, both Chell and Wallin models appeared not to be sufficiently conservative for LCF regime (prediction of “Case 2”); for other regimes (LUCF, LPUCF, LTUF and LPTUF) they appeared to be sufficiently conservative (in almost all cases). Based on the results of the tests, Wallin model was selected for implementation into the RPV integrity evaluation procedure, but simultaneously a decision was adopted to decrease its predictions when the “Case 2” is predicted: instead of predicting some surplus (15% of virgin KIC) above the value of KWPS, only value of KWPS (without any surplus) is predicted. This measure enhances conservativeness of the Wallin model to a sufficient level: the performed WPS experiments then well confirm the Wallin model predictions decreased in this manner. Taking 90% of the value of KWPS represents an additional margin implemented currently in the WPS methodology.

Deterministic and Probabilistic PTS Study for a Reactor Pressure Vessel Considering Plume Cooling Effects

2016 ◽  
Author(s):  
Guian Qian ◽  
Markus Niffenegger ◽  
Medhat Sharabi ◽  
Nathan Lafferty
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Cooling Effect ◽  
Integrity Assessment ◽  
Cooling Effects ◽  
Reactor Pressure ◽  
Element Method

A reactor pressure vessel (RPV) is assumed to be subjected to pressurized thermal shocks (PTSs) as a result of the emergency cooling water injected during a loss-of-coolant accident (LOCA). The cooling flow is not homogeneous but typically in a plume shape (stripe cooling) flowing from the cold leg through the inlet nozzles. This paper aims to analyze the non-uniform cooling effect on the RPV integrity. In this paper, both deterministic and probabilistic methods are used to analyze the integrity of a model RPV subjected to PTS. RELAP5, GRS-MIX, CFD and other semi-analytical methods are used to analyze the transient with and without considering plume cooling effect. Finite element method (FEM), extended finite element method (XFEM) and weight function method are used to calculate KI of the postulated cracks. The FAVOR code is used to calculate the conditional probabilities for crack initiation and failure of the RPV considering different crack distributions. KI based on CFD input is the highest, followed by that based on reference transient, GRS-MIX and RELAP5. Peak KI of the cracks inside the plume increases about 33% compared with that outside. According to the maximum criteria, the maximum allowed RTNDT are 56.9 °C, 90.2 °C, 98 °C, 115.7 °C and 136.2 °C for the crack in the nozzle region based on CFD transient, the cracks in the ring region based on the CFD, reference data, GRS-MIX and RELAP5 calculated transient, respectively. These values are 36 °C, 68.5 °C, 73 °C, 81 °C and 104 °C according to the tangent criteria. The conditional probability inside the plume is more than nine orders of magnitude higher than outside the plume. Considering plume cooling effects increases the total failure frequency by 1–2 orders of magnitude. In order to be conservative, it is necessary to consider the plume effect in the integrity assessment.

Probabilistic Pressurized Thermal Shocks Analyses for a Reactor Pressure Vessel

2015 ◽  
Vol 137 (6) ◽  
Author(s):  
Guian Qian ◽  
Markus Niffenegger
Keyword(s):  
Aspect Ratio ◽  
Pressure Vessel ◽  
Crack Depth ◽  
Crack Density ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Critical Crack ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

Both deterministic and probabilistic methods are used to assess the integrity of a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code is applied to calculate the probabilities for crack initiation and failure of the RPV subjected to two transients, by considering crack distributions based on cracks observed in the Shoreham and pressure vessel research user facility (PVRUF) RPVs. The crack parameters, i.e., crack density, depth, aspect ratio, orientation, and location are assumed as random variables following different distributions. KI of the cracks with the same depth increases with its aspect ratio. Both KI and KIc at the crack tip increase with crack depth, which is the reason why a deeper crack does not necessarily lead to a higher failure probability. The underclad crack is the most critical crack and the deeper crack is the least critical one in this study. Considering uncertainties of the transients results in higher failure probabilities.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

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