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.

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.

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.

Application of the Master Curve Approach for Abnormal Material Conditions

2007 ◽  
Author(s):  
Tapio Planman ◽  
William Server ◽  
Kim Wallin ◽  
Stan Rosinski
Keyword(s):  
Fracture Toughness ◽  
Lower Bound ◽  
Master Curve ◽  
Pressure Vessels ◽  
Data Sets ◽  
Inhomogeneous Materials ◽  
Data Set ◽  
Material Inhomogeneity ◽  
Fracture Modes ◽  
Material Conditions

The range of applicability of Master Curve testing Standard ASTM E 1921 is limited to macroscopically homogeneous steels with “uniform tensile and toughness properties”. A majority of structural steels appear to satisfy this requirement by exhibiting fracture toughness data which comply with the assumed KJc vs. temperature dependence and scatter within the specified validity area. As indicated in ASTM E 1921 a criterion for material macroscopic inhomogeneity is often applied using the 2% lower bound (possibly also the 98% upper bound). Data falling below this 2% lower-limit curve may be an indication of material inhomogeneity or susceptibility to grain boundary fracture. When this situation occurs, it is recommended to analyze the material with the so-called SINTAP procedure, which is intended for randomly inhomogeneous materials to assure a conservative lower-bound estimate. When a data set distinctly consists of two or more different data populations instead of one (due to variation of irradiation dose or specimen extraction depth, for instance) adoption of a bimodal (or a multimodal) Master Curve model is generally appropriate. These modal models provide information if the deviation of distributions is statistically significant or if different distributions truly exist for values of reference transition temperature, T0, characteristic of separate data populations. In the case of data sets representing thick-walled structures (i.e., reactor pressure vessels), indications of abnormal fracture toughness data can be encountered such that material inhomogeneity or fracture modes other than pure cleavage should be suspected. A state-of-the-art review for extended, non-standard Master Curve data and techniques highlights limits of applicability in situations where the basic ASTM E 1921 procedure is not appropriate for material homogeneity or different fracture modes.

IAEA Coordinated Research Project on Master Curve Approach to Monitor Fracture Toughness of RPV Steels: Effects of Bias, Constraint, and Geometry

2007 ◽  
Author(s):  
Randy K. Nanstad ◽  
Marc Scibetta
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Pressure Vessels ◽  
Specimen Size ◽  
Nuclear Industry ◽  
Reference Temperature ◽  
Bias Effect ◽  
Research Project ◽  
Energy Agency ◽  
Reactor Pressure

There is strong interest from the nuclear industry to use the precracked Charpy single-edge notched bend, SE(B), specimen (PCVN) to enable determination of the reference temperature, T0, with reactor pressure vessel surveillance specimens. Unfortunately, for many different ferritic steels, tests with the PCVN specimen (10×10×55 mm) have resulted in T0 temperatures up to 25°C lower than T0 values obtained using data from 25-mm thick compact specimens [1TC(T)]. This difference in T0 reference temperature has often been designated a specimen bias effect, and the primary focus for explaining this effect is loss of constraint in the PCVN specimen. The International Atomic Energy Agency has developed a three-part coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water reactor pressure vessels. One part of the CRP is focused on the issue of test specimen geometry effects, with emphasis on the PCVN bias. Participating organizations for this part of the CRP are performing fracture toughness testing of various steels, including the reference steel JRQ (A533-B-1) often used for IAEA studies, with various types of specimens under various conditions. Additionally, many of the participants are taking part in a round robin exercise on finite element modeling of the PCVN specimen. Some preliminary results from fracture toughness tests are compared with regard to effects of specimen size and type on the reference temperature T0. In agreement with a number of published results, the results do generally show lower values of T0 from the PCVN specimen compared with the compact and larger bend specimens. They also clearly show higher apparent fracture toughness for the shallow crack compared with the deep crack configuration. Moreover, the SE(B) specimens exhibit a tendency for decreasing T0 with decreasing specimen size (thickness and/or remaining ligament). Additionally, as shown in previous CRPs, the results also exhibit a dependence on test temperature. Following completion of all testing, the results will be evaluated relative to existing proposed models with a view towards developing an understanding of the reasons for the observed differences.

Large SA-508 Class 2 Nozzle Forging Near-Surface Fracture Toughness

2017 ◽  
Author(s):  
J. Brian Hall ◽  
Benjamin E. Mays ◽  
Matthew DeVan
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Lower Bound ◽  
Master Curve ◽  
Impact Test ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Fracture Toughness Test ◽  
Test Technique ◽  
Near Surface

The current approach in evaluating the Pressurized Water Reactor (PWR) inlet and outlet nozzle corner regions with respect to plant heat-up and cool-down pressure-temperature limit curves contains a number of conservatisms. These conservatisms include postulation of a large ¼ thickness flaw at the nozzle corner region and use of RTNDT (reference nil-ductility temperature) or an estimation of RTNDT. The paper herein discusses generic fracture toughness of nozzle forging material SA-508 Class 2 for use with postulated smaller surface flaws in developing pressure-temperature limit curves for nozzle corners for nuclear power plant operations. ASME Appendix G uses the lower bound KIC curve, which has inherent margin since RTNDT is a conservative method for locating the KIC curve. RTNDT is based on the drop weight test, which is a crack arrest transition temperature measurement, and the Charpy impact test, which is a blunt notch impact test. These data are conservatively bounded by the KIC curve, which is a lower bound crack initiation toughness curve. In contrast, the master curve method is based on an initiation transition temperature fracture toughness test technique per ASTM E1921. The master curve index temperature (T0) provides a more accurate measure of the material fracture toughness than KIC indexed with RTNDT. Since many of the nuclear pressure vessels were fabricated to ASME Code editions prior to 1972, RTNDT was not measured for the nozzles. In many cases, RTNDT has been estimated. Therefore, for this work, the fracture toughness was generically established based on conservative T0 measurements of 22 representative forgings with a margin of two standard deviations to ensure a conservative lower bound toughness using ASME Appendix G, G-2110. The properties of a forging are better near the surface due to the faster cooling rate during heat treatment. The difference in reactor pressure vessel fracture toughness was established for forgings near the surface at the postulated flaw location as allowed by ASME Section III, NB-2223.2 relative to the traditional ¼ thickness location. The near-surface forging toughness was conservatively determined through evaluation of 31 near-surface and approximate ¼ thickness location fracture toughness measurements.

Complex Monitoring Programme for WWER Reactor Pressure Vessels Lifetime Assessment in the Czech Republic

2008 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka ◽  
Petr Novosad ◽  
Jiri Brynda
Keyword(s):  
Czech Republic ◽  
Weld Metal ◽  
Heat Affected Zone ◽  
Neutron Fluence ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
The Czech Republic ◽  
Complex Monitoring ◽  
Metal Weld ◽  
Reactor Pressure

Lifetime of reactor pressure vessels practically depends on a level of degradation of RPV material properties during operation. The most important degradating mechanism of RPV materials is usually radiation damage, characterized by values on neutron fluence on one side and radiation embrittlement of RPV materials on the second side. WWER reactor pressure vessels in the Czech Republic are a subject of a very thorough and complex monitoring program, that includes: • Standard material surveillance program containing of WWER-440 RPV materials — base metal, weld metal, heat affected zone, but irradiated with high lead factor (13 to 18), • Supplementary surveillance program of WWER-440 RPV materials, including additionally austenitic cladding materials, IAEA reference material JRQ irradiated with low lead factor (2 to 3) with parts subjected to annealing and re-irradiation after annealing, • Modified surveillance program of WWER-1000 RPV materials — base metal, weld metal, heat affected zone, cladding materials, IAEA reference JRQ material irradiated in low lead factor (2 to 3) near RPV inner beltline region, • Integrated surveillance specimen program for WWER-1000 reactor including materials from NPP Temelin (Czech Republic), Belene (Bulgaria), Kalinin (Russia) and Ukranian NPPs, • Continous exvessel monitoring of neutron fluence on outer RPV surface for both WWER-440 and WWER-1000 plants, • Neutron fluence determination on inner RPV surface (austenitic cladding) using special technique for removal of specimens from cladding for Nb activity measurements, • Ex-vessel temperature measurements during RPV operation. All these programs serve for precision of operation conditions and determination of degradation of RPV materials for RPV integrity and lifetime assessment.

Master Curve and J-R Fracture Toughness of SA508/SA533-B-1 Weld and HAZ

2013 ◽  
Author(s):  
Randy K. Nanstad ◽  
Xiang Chen ◽  
Mikhail A. Sokolov ◽  
Barry H. Rabin ◽  
Ying Yang
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Master Curve ◽  
Potential Drop ◽  
Material Behavior ◽  
Normalization Procedure ◽  
Threshold Temperatures ◽  
Class 1 ◽  
Large Heat ◽  
Grade 3

A large heat of low-alloy steel that met both specifications for SA508 Grade 3 Class1 forging steel and SA533 Type B Class 1 plate steel (A508/A533) was procured and used to fabricate a submerged-arc weldment for potential application in high temperature gas-cooled reactors. Compact specimens, 1TC(T), were machined from the weld metal and from the heat-affected-zone (HAZ) of the weldment. Tests of both materials were performed to obtain the fracture toughness reference temperature, To, using the Master Curve procedure of ASTM E-1921, and J-R curves to evaluate material behavior at various threshold temperatures in Code Case N-499-2 (2001) of the ASME Boiler and Pressure Vessel Code. Tests were performed at various temperatures up to 593°C. Unloading compliance was the primary technique used, although dc-potential drop was also monitored during the tests, and the normalization procedure of E1820 was used to compare the results from each procedure. Moreover, many tests at the highest temperatures were performed with no unloading and the normalization procedure provided in E1820 was used to analyze the load-displacement measurements. The fracture toughness for the HAZ is superior to that of the weld metal both in terms of transition temperature and ductile fracture toughness.

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.

Prediction of Fracture Toughness for WWER RPV Integrity Assessment on the Basis of the Unified Curve Approach and Surveillance Specimens Testing

2009 ◽  
Author(s):  
Boris Margolin ◽  
Victoria Shvetsova ◽  
Alexander Gulenko ◽  
Valentin Fomenko
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Present Report ◽  
Temperature Shift ◽  
Curve Shape ◽  
Integrity Assessment ◽  
Master Curve Method ◽  
Curve Method ◽  
Unified Curve ◽  
Surveillance Specimens

For construction of the fracture toughness temperature curve that may be used for WWER RPV integrity assessment on the basis of tests of cracked surveillance specimens, the issues have to be solved as follows. First of all, it is important to determine how fracture toughness varies as a function of temperature, and how the fracture toughness vs. temperature dependence, KJC(T), changes with in-service material degradation due to neutron irradiation. These variations of KJC(T) curve are known to be the shift of KJC(T) curve to higher temperature range and change in the KJC(T) curve shape. At present, two advanced engineering methods are known that allow the prediction of KJC(T) curve on the basis of small-size fracture toughness specimens (for example, pre-cracked Charpy specimens), namely, the Master Curve and the Unified Curve methods. Procedures of test result treatment for the Master Curve and the Unified Curve are very similar. The Master Curve method uses the lateral temperature shift condition and, therefore, does not describe possible change in the KJC(T) curve shape. The Unified Curve method has an advantage as compared with the Master Curve as the Unified Curve describes a variation of the KJC(T) curve shape when degree of embrittlement increases. This advantage becomes important for RPV integrity assessment when the reference KJC(T) curve is recalculated to the crack front length of the postulated flaw that is considerable larger than thickness of surveillance specimens. Application of the KJC(T) curve determined from test results of cracked surveillance specimens to RPV integrity assessment requires also to introduce some margins. These margins have to take into account the type and number of tested specimens and the uncertainty connected with spatial non-homogeneity of RPV materials. Indeed, there is sufficient number of experimental data showing variability in fracture toughness for various parts of RPV. Therefore, situation is possible when the material properties near the postulated flaw will be worse than the properties of surveillance specimens. In the present report, advanced approaches are considered for prediction of fracture toughness for WWER RPV integrity assessment that allow one: • to construct the KJC(T) curve for irradiated RPV steels with any degree of embrittlement; • to provide transferability of fracture toughness data from cracked surveillance specimens to calculation of resistance to brittle fracture of RPV with a postulated flaw.

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

2014 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Analysis Method ◽  
Integrity Assessment ◽  
The U.S

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration. In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

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