French Four-Loop PWR 1300 MWe Reactor Pressure Vessel: First Results for Integrity Assessment and Life Manangement

2003 ◽  
Author(s):  
Georges Bezdikian ◽  
Franc¸oise Ternon-Morin ◽  
Henriette Churier-Bossennec ◽  
Dominique Moinereau ◽  
Alain Martin
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Surveillance Program ◽  
Integrity Assessment ◽  
First Results ◽  
Safety Studies ◽  
Reactor Pressure

The process used by the French utility, concerning the Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs 3-loop Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, was engaged several years ago and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection for beyond 40 years. Since 2000, in the continuity of this results, the studies was carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained shows the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering major parameters particularly the severity of the transient. This approach, is based on specific mechanical safety studies on the 1300 MWe RPV, to demonstrate the absence of risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 4-loop RPV RTNDT during the lifetime in operation, 2 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of the brittle fracture. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • better knowledge of the vessel material properties, including the effect of radiation, • more precise assessment of the fluence and neutronic calculations, • the NDE inspection program based on the inspection of the vessel wall, with a special NDE tool. The principal actions conducted during recent years are: • the optimisation of the fuel management and the new development for the fluence evaluation, • the data gathered from radiation specimen capsules, removed from the vessels (4loop reactor), within the framework of the radiation surveillance program, and the thermal-hydraulic-mechanical calculations based on finite element thermal-hydraulic computations and three dimensional elastic-plastic mechanical analyses.

RPV Integrity Assessment: Comparison Between 3 Loops and 4 Loops Studies

2004 ◽  
Author(s):  
Georges Bezdikian ◽  
Dominique Moinereau ◽  
Claude Faidy
Keyword(s):  
Brittle Fracture ◽  
Nuclear Power ◽  
Power Plants ◽  
Electric Energy ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
Integrity Assessment ◽  
Safety Studies ◽  
Comparison Of The Results ◽  
Reactor Pressure

For the French utility (Electricite de France–EDF), Nuclear Energy represents 75% of generation of the total electric energy in France. Total nuclear electricity were generated mainly from Nuclear Power plants stations, 34 PWR NPPs 3-loop Reactors- 900 MWe, 20 PWR NPPs 4-loop Reactors- 1300 MWe and 4 PWR NPPs 4-loop Reactors- 1450 MWe. The 3-loop Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection beyond 40 years. Since 2000, in the continuity of these results, the studies were carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained show the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and other major parameters. This approach is based on specific mechanical safety studies on the RPV to demonstrate the absence of the risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 3-loop and 4-loop RPV, 2 - RTNDT during the lifetime in operation, 3 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of brittle fracture for all of 3-loop and 4-loop RPV. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • more precise assessment of the fluence calculations, • better knowledge of the vessel material properties, including the effect of radiation, • the NDE inspection program on the core zone. The comparison of the results are developed in this paper: • for the fluence evaluation and the optimisation of the fuel management, • the data gathered from radiation specimen capsules, removed from the vessels (radiation surveillance program), • and the thermal-hydraulic and mechanical calculations based on finite element thermal-hydraulic and 3D elastic-plastic mechanical computations.

Using a Probabilistic Approach in the Brittle Fracture Deterministic Integrity Assessment of a Nuclear Reactor Pressure Vessel

2011 ◽  
Author(s):  
Romain Beaufils ◽  
Eric Meister ◽  
Emmanuel Ardillon
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Probabilistic Method ◽  
Probabilistic Approach ◽  
Reactor Pressure Vessel ◽  
Deterministic Method ◽  
Integrity Assessment ◽  
Reactor Pressure

This work deals with the possibility of the life extension of nuclear power plants in France. The aim is to justify the resistance of the pressure vessel, which is non-replaceable. The brittle fracture deterministic integrity assessment of the nuclear Reactor Pressure Vessel (RPV) is based on the analysis of a flaw under the austenitic cladding of the RPV. The demonstration of the RPV resistance is controlled by the regulations. It is proposed here to use a probabilistic method by propagating uncertainties into the deterministic mechanical model in order to quantify conservatism of the deterministic method. The regulatory requirements must be respected and the purpose of the work presented here is thus to link the probabilistic result to the deterministic method.

Ways for Extending the Reactor Pressure Vessel Lifetime

2008 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Low Leakage ◽  
Vessel Integrity ◽  
Direct Use ◽  
Reactor Pressure ◽  
Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

MAI Benchmark Campaign of International Software for Reactor Pressure Vessel Integrity Assessment

2014 ◽  
Author(s):  
Emilie Dautreme ◽  
Emmanuel Remy ◽  
Roman Sueur ◽  
Jean-Philippe Fontes ◽  
Karine Aubert ◽  
...  
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Software Tool ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Benchmark Study ◽  
Reactor Pressure

Nuclear Reactor Pressure Vessel (RPV) integrity is a major issue concerning plant safety and this component is one of the few within a Pressurized Water Reactor (PWR) whose replacement is not considered as feasible. To ensure that adequate margins against failure are maintained throughout the vessel service life, research engineers have developed and applied computational tools to study and assess the probability of pressure vessel failure during operating and postulated loads. The Materials Ageing Institute (MAI) sponsored a benchmark study to compare the results from software developed in France, Japan and the United States to compute the probability of flaw initiation in reactor pressure vessels. This benchmark study was performed to assess the similarities and differences in the software and to identify the sources of any differences that were found. Participants in this work included researchers from EDF in France, CRIEPI in Japan and EPRI in the United States, with each organization using the probabilistic software tool that had been developed in their country. An incremental approach, beginning with deterministic comparisons and ending by assessing Conditional Probability of crack Initiation (CPI), provided confirmation of the good agreement between the results obtained from the software used in this benchmark study. This conclusion strengthens the confidence in these probabilistic fracture mechanics tools and improves understanding of the fundamental computational procedures and algorithms.

scholarly journals Influence of preliminary thermomechanical load on the integrity of the vessel of Ukrainian nuclear reactors in case of accidents of "large" and "small" leak

2021 ◽  
Author(s):  
Oleksii Ishchenko
Keyword(s):  
Brittle Fracture ◽  
Pressure Vessel ◽  
Fracture Resistance ◽  
Emergency Situation ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Brittle Strength ◽  
Brittle Fracture Resistance ◽  
Small Leak ◽  
Reactor Pressure

In works of extending lifetime of WWER-type reactors, it is necessary to obtain the brittle fracture resistance (BF) of the reactor pressure vessel (RPV). Implementation calculations for brittle fracture resistance is regulated by the technical PNAE of Ukraine. The objective of these calculations is to prevent catastrophic brittle destruction of the reactor pressure vessel, pipelines and pressure vessels from the existence crack-like defects for all operating regimes, including emergency situations (ES). The paper considers the most dangerous postulated emergency situation in operation is "large" and "small" leak in the RPV NPP. Calculations with Warm Pre-Stressing effect (WPS) of the RPV for the most dangerous scenarios have been presented, and an assessment of the brittle strength of RPV NPP is taking into account with WPS. The results of studies with factor of brittle strength safety are also presented without taking into account the Warm Pre-Stressing, comparing with the existing method for accounting this type of load.

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.

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

2014 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Tadashi Watanabe ◽  
Yutaka Nishiyama
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Coolant Water ◽  
Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

2013 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Detailed Comparison ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Asme Code ◽  
Vessel Integrity ◽  
Degradation Of Materials ◽  
Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

2013 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Detailed Comparison ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Asme Code ◽  
Vessel Integrity ◽  
Degradation Of Materials ◽  
Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

Features and Results of Several Programs of Radiation-Damage Surveillance of Power Reactor Pressure Vessels

1967 ◽  
Vol 89 (1) ◽  
pp. 221-232 ◽  
Author(s):  
Charles Z. Serpan ◽  
L. E. Steele ◽  
J. R. Hawthorne
Keyword(s):  
Radiation Damage ◽  
Pressure Vessel ◽  
Power Reactor ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Surveillance Program ◽  
Surveillance Programs ◽  
Meaning And Purpose ◽  
Reactor Pressure

The meaning and purpose of reactor pressure vessel surveillance is briefly discussed. Features of the surveillance programs in the Yankee, Big Rock Point, SM-1, and SM-1A reactors are briefly described along with results of testing metallurgical specimens from these programs. Additionally, the surveillance program to be effected in the Army MH-1A reactor is described. Certain problems which have occurred in the course of these programs are discussed as well as the proposed ASTM recommendations for radiation-damage surveillance programs. The value of these programs to reactor operators is reviewed with relation to the results obtained to date.

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