Validation of RTT0 for German Reactor Pressure Vessel Steels

2008 ◽  
Vol 130 (3) ◽  
Author(s):  
Dieter Siegele ◽  
Elisabeth Keim ◽  
Gerhard Nagel
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Data Sets ◽  
Asme Code ◽  
Definition Of ◽  
Size Adjustment ◽  
Reactor Pressure

For the introduction of the new reference temperature RTT0 of the ASME Code Cases N-629 and N-631 into the German Standard KTA 3201.2, the applicability of RTT0 was validated by the reevaluation of the existing fracture toughness database of German reactor pressure vessel. (RPV) steels including unirradiated and irradiated base materials and weld metal data. The test temperatures of the database were standardized to the reference temperature T0 of the master curve of the data sets and the database was compared with the ASME KIC-curve as adjusted by RTT0. The KIC-curve adjusted by RTT0 enveloped both the 1T-size adjusted database and also the as-measured database, corresponding to the definition of RTT0. Thus, the results also prove the validity of the KIC(RTT0)-curve for allowable flaw sizes and up to the crack length spectrum of the ASME KIC-database without size adjustment of T0. The results of both investigations confirmed the validity of RTT0 for German RPV steels. The majority of existing fracture toughness data are based on KIC-values. More recent data are (KJC) related to the issuing of ASTM E 1921 in 1997 and to the success of the master curve-based T0 approach. Therefore, the possible difference between T0 determined from KJC and from KIC was investigated with available databases for RPV steels. The comparison of T0(KJC) and T0(KIC) showed a 1:1 correlation proving the equivalence of KJC and KIC in the determination of T0.

Validation of RTTo for German Reactor Pressure Vessel Steels

2005 ◽  
Author(s):  
Elisabeth Keim ◽  
Dieter Siegele ◽  
Gerhard Nagel
Keyword(s):  
Data Base ◽  
Base Material ◽  
Measured Data ◽  
Reference Temperature ◽  
Data Sets ◽  
Pressure Vessel Steels ◽  
Asme Code ◽  
Definition Of ◽  
Size Adjustment ◽  
Reactor Pressure

For the introduction of the new reference temperature RTTo of the ASME Code Cases N-629 and N-631 into the German KTA rules the equation RTTo = T0 + 19,4 °C (35 °F) was validated by the re-evaluation of the existing fracture toughness data base of German RPV steels, including unirradiated and irradiated base material and weld metal data. The test temperatures of the data base were standardized to the reference temperature T0 of the Master Curve of the data sets and the data base was compared with the ASME KIc-curve as adjusted by RTTo. The KIc-curve adjusted by RTTo enveloped both, the 1T-size adjusted data base and also the as measured data base, corresponding with the definition of RTTo. Thus the results also prove the validity of the KIc(RTTo)-curve for allowable flaw sizes and up to the crack length spectrum of the ASME KIC-data base without size adjustment of T0. The results of both investigations confirmed the validity of RTTo for German RPV steels.

Application of Master Curve Data for Reactor Vessel Steels

2003 ◽  
Author(s):  
William L. Server ◽  
Timothy J. Griesbach ◽  
Stan T. Rosinski
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Material Behavior ◽  
Data Sets ◽  
Master Curve Method ◽  
Curve Method ◽  
Curve Fracture ◽  
Reactor Pressure

The Master Curve method has been developed to determine fracture toughness of a specific material in the brittle-to-ductile transition range. This method is technically more descriptive of actual material behavior and accounts for the statistical nature of fracture toughness properties as an alternative to the current ASME Code reference toughness curves. The Master Curve method uses a single temperature, To, as an index of the Master Curve fracture toughness transition temperature. This method has been successfully applied to numerous fracture toughness data sets of pressure vessel steels contained in the Master Curve database, including the beltline materials for the Kewaunee reactor pressure vessel. The database currently contains over 5,500 toughness data records for vessel weld, plate and forging materials, and it is currently being updated to include more recent fracture toughness data. Application of Master Curve fracture toughness data to reactor pressure vessel (RPV) integrity evaluations requires some assumptions relative to the degree of constraint in the fracture toughness test specimens versus the actual assumed RPV flaw. An excessive degree of conservatism can be introduced if the constraint levels are substantially different. In performing a Master Curve evaluation, the analysis may be restricted by the type of fracture toughness data available. Any excess conservatism should be appropriately considered when the overall safety margin is applied. For example, the precracked Charpy three-point bend specimen actually has some advantages over the compact tension specimen when the application involves a shallow surface flaw in a RPV wall. This paper analyzes some key fracture toughness results from several weld data sets containing both unirradiated and irradiated data to evaluate constraint effects in fracture toughness and pre-cracked Charpy specimens. The evaluated To values were compared to determine if there is any difference in bias from specimen geometry between the unirradiated and irradiated data.

Alternative Reference Temperature Based on Master Curve Approach in Japanese Reactor Pressure Vessel Steel

2013 ◽  
Author(s):  
Takatoshi Hirota ◽  
Takashi Hirano ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Reference Temperature ◽  
Ferritic Steels ◽  
Pressure Vessel Steel ◽  
Reactor Pressure

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011, “Test Method for Determination of Reference Temperature, To, of Ferritic Steels” was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature, To can be determined in accordance with this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels including base metals and weld metals, the method for determination of the alternative reference temperature RTTo based on Master Curve reference temperature To was statistically examined, so that RTTo has an equivalent safety margin to the conventional RTNDT. Through the statistical treatment, the alternative reference temperature RTTo was proposed as the following equation; RTTo = To + CMC + 2σTo. This method is applicable to the Japan Electric Association Code JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” as an option item.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

2009 ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Root Region ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do 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 evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

Estimation of Master Curve Based RTTo Reference Temperature From CVN Data

2005 ◽  
Author(s):  
Kim R. W. Wallin ◽  
Gerhard Nagel ◽  
Elisabeth Keim ◽  
Dieter Siegele
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Master Curve ◽  
Simple Relation ◽  
Systematic Evaluation ◽  
Data Sets ◽  
Irradiation Embrittlement ◽  
Pressure Vessel Steels ◽  
Asme Code

The ASME code cases N-629 and N-631 permits the use of a Master Curve-based index temperature (RTTo ≡ T0 + 19.4°C) as an alternative to traditional RTNDT-based methods of positioning the ASME KIc, and KIR curves. This approach was adopted to enable use of Master Curve technology without requiring the wholesale changes to the structure of the ASME Code that would be needed to use all aspects of Master Curve technology. For the brittle failure analysis considering irradiation embrittlement additionally a procedure to predict the adjustment of fracture toughness for EOL from irradiation surveillance results must be available as by NRC R.G. 1.99 Rev. 2 e.g.: ART = Initial RTNDT + ΔRTNDT + Margin. The conservatism of this procedure when RTNDT is replaced by RTTo is investigated for western nuclear grade pressure vessel steels and their welds. Based on a systematic evaluation of nearly 100 different irradiated material data sets, a simple relation between RTToirr, RTToref and ΔT41JRG is proposed. The relation makes use of the R.G. 1.99 Rev. 2 and enables the minimizing of margins, necessary for conventional correlations based on temperature shifts. As an example, the method is used to assess the RTTo as a function of fluence for several German pressure vessel steels and corresponding welds. It is shown that the method is robust and well suited for codification.

Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

2008 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Takashi Hirano ◽  
Seiji Asada ◽  
Ryoichi Saeki ◽  
Naoki Miura ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steels ◽  
The World ◽  
Reactor Pressure ◽  
Plant Components

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

Evaluation of Fracture Toughness of a SA580 Gr. 3 Reactor Pressure Vessel Using a Bimodal Master Curve Approach

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Jong-Min Kim ◽  
Seok-Min Hong ◽  
Min-Chul Kim ◽  
Bong-Sang Lee
Keyword(s):  
Fracture Toughness ◽  
Cooling Rate ◽  
Pressure Vessel ◽  
Master Curve ◽  
Random Parameter ◽  
Random Inhomogeneity ◽  
Reactor Pressure Vessel ◽  
Inhomogeneous Materials ◽  
Material Inhomogeneity ◽  
Reactor Pressure

Abstract The standard master curve (MC) approach has a major limitation in that it is only applicable to homogeneous datasets. In nature, steels are macroscopically inhomogeneous. Reactor pressure vessel (RPV) steel has different fracture toughness with varying distance from the inner surface of the wall due to the higher cooling rate at the surface (deterministic material inhomogeneity). On the other hand, the T0 value itself behaves like a random parameter when the datasets have large scatter because the datasets are for several different materials (random inhomogeneity). In this paper, four regions, the surface, 1/8 T, 1/4 T, and 1/2 T, were considered for fracture toughness specimens of Korean Standard Nuclear Plant (KSNP) SA508 Gr. 3 steel to provide information on deterministic material inhomogeneity and random inhomogeneity effects. Fracture toughness tests were carried out for the four regions at three test temperatures in the transition region and the microstructure of each region was analyzed. The amount of upper bainite increased toward the center, which has a lower cooling rate; therefore, the center has lower fracture toughness than the surface so reference temperature (T0) is higher. The fracture toughness was evaluated using the bimodal master curve (BMC) approach. The results of the BMC analyses were compared with those obtained via a conventional master curve analyses. The results indicate that the bimodal master approach considering inhomogeneous materials provides a better description of scatter in the fracture toughness data than a conventional master curve analysis does.

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.

Estimation of Master Curve Based RTTO Reference Temperature From CVN Data

2006 ◽  
Vol 129 (3) ◽  
pp. 420-425
Author(s):  
Kim R. W. Wallin ◽  
Gerhard Nagel ◽  
Elisabeth Keim ◽  
Dieter Siegele
Keyword(s):  
Pressure Vessel ◽  
Master Curve ◽  
Simple Relation ◽  
Reference Temperature ◽  
Systematic Evaluation ◽  
Data Sets ◽  
Irradiation Embrittlement ◽  
Pressure Vessel Steels ◽  
Asme Code ◽  
Additional Procedure

The ASME code cases N-629 and N-631 permit the use of a master curve-based index temperature (RTTo≡T0+19.4°C) as an alternative to traditional RTNDT-based methods of positioning the ASME KIC and KIR curves. This approach was adopted to enable the use of master curve technology without requiring the wholesale changes to the structure of the ASME code that would be needed to use all aspects of master curve technology. For the brittle failure analysis considering irradiation embrittlement an additional procedure to predict the adjustment of fracture toughness for end of life (EOL) from irradiation surveillance results must be available as by NRC R.G. 1.99 Rev. 2, e.g., the adjusted reference temperature is defined as ART=initialRTNDT+ΔRTNDT+margin. The conservatism of this procedure when RTNDT is replaced by RTTo is investigated for western nuclear grade pressure vessel steels and their welds. Based on a systematic evaluation of nearly 100 different irradiated material data sets, a simple relation between RTToirr, RTToref, and ΔT41JRG is proposed. The relation makes use of the R.G. 1.99 Rev. 2 and enables the minimizing of margins, necessary for conventional correlations based on temperature shifts. As an example, the method is used to assess the RTTo as a function of fluence for several German pressure vessel steels and corresponding welds. It is shown that the method is robust and well suited for codification.

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