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

2010 ◽  
pp. 115-115-14
Author(s):  
Hans -Werner Viehrig ◽  
Jan Schuhknecht ◽  
Udo Rindelhardt ◽  
Frank -Peter Weiss
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Welding Seam ◽  
Reactor Pressure

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.

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

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Power Plants ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPPs) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterization. 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 focused on the characterization 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 American Society for Testing of Materials (ASTM) Test Standard E1921–08 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 (MC) 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 behavior. 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 subsize 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-V transition temperature TT41J estimated with results of subsize specimens after the recovery annealing was confirmed by the testing of standard Charpy-V-notch specimens. The evaluated 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. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during two campaign operations can be assumed to be low for the weld and base metal.

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

2009 ◽  
Vol 6 (5) ◽  
pp. 101925 ◽  
Author(s):  
Hans-Werner Viehrig ◽  
Jan Schuhknecht ◽  
Udo Rindelhardt ◽  
Frank-Peter Weiss ◽  
Jeremy T. Busby ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Welding Seam ◽  
Reactor Pressure

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure

The effects of thermal stress on the crack propagation in AP1000 reactor pressure vessel

2020 ◽  
Vol 110 ◽  
pp. 102798
Author(s):  
KaiTai Liu ◽  
Mei Huang ◽  
JunJie Lin ◽  
HaiPeng Jiang ◽  
BoXue Wang ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Crack Propagation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Ap1000 Reactor ◽  
Reactor Pressure

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

The ageing behavior and the correlation between hardness and coercivity of Cu-rich reactor pressure vessel model steels

2021 ◽  
Vol 527 ◽  
pp. 167698
Author(s):  
Xuejiao Wang ◽  
Wenjiang Qiang ◽  
Guogang Shu ◽  
Junwei Qiao ◽  
Yucheng Wu
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure
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