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.

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.

Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens

2018 ◽  
Author(s):  
Michael R. Ickes ◽  
J. Brian Hall ◽  
Robert G. Carter
Keyword(s):  
Fracture Toughness ◽  
Direct Measurement ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Weld Metals ◽  
Surveillance Programs ◽  
Test Reactor ◽  
Reactor Pressure

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

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.

Application of Master Curve Approach to Surveillance Test Data for WWER-1000 Reactor Pressure Vessels

2017 ◽  
Author(s):  
Volodymyr M. Revka ◽  
Liudmyla I. Chyrko
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Test Data ◽  
Master Curve ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Transition Curve ◽  
Surveillance Program ◽  
Data Set ◽  
Reactor Pressure

An important issue in the safety operation of WWER-1000 type reactor is a decrease in fracture toughness for reactor pressure vessel steels due to neutron irradiation. This effect for RPV metal is known as radiation embrittlement. The radiation induced temperature shift of the fracture toughness transition curve is considered as a measure of the embrittlement rate. The Charpy impact and fracture toughness specimens are included in the surveillance program for an assessment of changes in fracture toughness of RPV materials. The present analysis is based on a large data set which includes mostly experimental results for pre-cracked Charpy specimens from a WWER-1000 RPV surveillance program. A Master curve approach is applied to analyze the surveillance test data with respect to a shape of the fracture toughness transition curve and a scatter of KJC values. The RPV base and weld metal in unirradiated, thermally aged and irradiated conditions are considered in this study. The maximum shift in a reference temperature T0 due to irradiation is 107 degree Celsius. It is shown that the Master curve, 5 % and 95 % tolerance bounds describe adequately the temperature dependence and the statistical scatter of KJC values for WWER-1000 RPV steels both in unirradiated condition and after irradiation up to design as well as long term operation neutron fluence. Furthermore, a development of the Weibull plots for considered data sets is shown that the Weibull slope is close to the expected one of 4 on average. Finally, a comparison of the reference temperature T0 and a scatter of KJC values derived from the pre-cracked Charpy and 0,5T C(T) specimens of base and weld metal in unirradiated condition is done. The analysis has shown a significant discrepancy between the T0 values derived from the two different types of specimens for both RPV metals.

Effect of Plastic Constraint and Cladding on Semi-Elliptical Shaped Crack in Fracture Toughness Evaluation for a Reactor Pressure Vessel Steel

2021 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Yasuto Nagoshi ◽  
Kai Lu ◽  
Jinya Katsuyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Integrity Assessment ◽  
Plastic Constraint ◽  
Shaped Crack ◽  
Reactor Pressure

Abstract In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (KJc) should be higher than the stress intensity factor at the crack tip of a semi-elliptical shaped under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and KJc evaluation. In this study, we performed fracture toughness tests and finite element analyses to investigate the effect of plastic constraint and cladding on the semi-elliptical shaped crack in KJc evaluation. The apparent KJc value evaluated at the deepest point of the crack exceeded 5% fracture probability based on the Master Curve method estimated from C(T) specimens, and the conservativeness of the current integrity assessment method was confirmed. Few initiation sites were observed along the tip of semi-elliptical shaped crack other than the deepest point. The plastic constraint state was also analyzed along the crack tip, and it was found that the plastic constraint at the crack tip near the surface was lower than that for the deepest point. Moreover, it was quantitatively showed that the UCC decreased the plastic constraint. The local approach suggested higher KJc value for the UCC than that for the surface crack, reflecting the low constraint effect for the UCC.

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

2008 ◽  
Author(s):  
Udo Rindelhardt ◽  
Hans-Werner Viehrig ◽  
Joerg Konheiser ◽  
Jan Schuhknecht
Keyword(s):  
Fracture Toughness ◽  
Lower Bound ◽  
Weld Metal ◽  
Master Curve ◽  
Neutron Fluence ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Integrity Assessment ◽  
The Core ◽  
Inner Surface

Between 1973 and 1990 4 units of the Russian NPP type WWER-440/230 were operated in Greifswald (former GDR). The operation was stopped after the German reunification, because the units did not completely follow western nuclear safety standards. 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. This weld was annealed after 13 cycles and operated further for 2 cycles. Additionally, starting with cycle 11, dummy assemblies were inserted to reduce the neutron fluence in the RPV wall. Firstly this paper presents results of the 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 fluence increase is reduced to 1/6 at the position of the maximum fluence. The neutron fluence at the different circumferential welds is closely related to their distance to the core. 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 neutron fluences at the both other welds are 2 resp. 4 orders of magnitude smaller according to their distances to the core. It should be mentioned that in this cases the fluence gradient can be negative through the wall. The material investigations were done using a trepan from the circumferential core weld. Master Curve and Charpy V-notch testing were applied. Specimens from 7 locations through the thickness of the welding seam were tested. The reference temperature T0 was calculated with the measured fracture toughness values, KJc, at brittle failure of the specimen. Generally the KJc values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follows the course of the Master Curve. The KJc values show a remarkable scatter. In addition the MC SINTAP procedure was applied to determine T0SINTAP of the brittle fraction of the data set. There are remarkable differences between T0 and T0SINTAP indicating macroscopic inhomogeneous weld metal. The highest T0 was about 50°C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T0 at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material does not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size 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. It enables the determination of a reference temperature, RTT0 to index a lower bound fracture toughness curve. This curve agrees with the MC 5% fractile as specified in ASTM E1921-05. The measured KJc values are not enveloped by this lower bound curve. However, the VERLIFE lower bound curve indexed with the 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 dataset of measured KJc values has to be applied.

The Fracture Toughness Properties of China Manufactured Reactor Pressure Vessel Steels in Transition Temperature Range

2018 ◽  
Author(s):  
Kai Sun ◽  
Xiaoyong Wu ◽  
Guoyun Li ◽  
Bang Wen
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Temperature Range ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Over the past decade, many generation III pressurized water reactor power plants have been under construction in China. Most reactor pressure vessel steels for these plants construction are homemade. Historically, Charpy V-notch specimens are predominantly used to monitor the toughness of RPV steels. However, fracture toughness provides the quantitative predictions of the critical crack size and the allowable stress in structural integrity assessment. This paper evaluates the fracture toughness properties of China manufactured RPV steels directly measured in transition temperature range by using master curve method. Some specimens were irradiated in the High Flux Engineering Test Reactor. The influences of loading rate, test temperature, specimen configuration and neutron irradiation on T0 were also investigated. The experimental results show that China manufactured RPV steels exhibit good fracture toughness properties.

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.

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.

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