Study of the Beltline Weld and Base Metal of WWER-440 First Generation Reactor Pressure Vessel
Представлены результаты исследования материалов бандажного сварного шва и кольца
 основного металла сосуда давления первого блока реактора типа ВВЭР-440/230. Исследовали
 круглые вырезки материала (трепаны) после радиационного облучения с последующим отжигом и повторным облучением....
Saved in:
| Published in: | Проблемы прочности |
|---|---|
| Date: | 2010 |
| Main Authors: | Schuhknecht, J., Rindelhardt, U., Viehrig, H.W. |
| Format: | Article |
| Language: | English |
| Published: |
Інститут проблем міцності ім. Г.С. Писаренко НАН України
2010
|
| Subjects: | |
| Online Access: | https://nasplib.isofts.kiev.ua/handle/123456789/111649 |
| Tags: |
Add Tag
No Tags, Be the first to tag this record!
|
| Journal Title: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Cite this: | Study of the Beltline Weld and Base Metal of WWER-440 First Generation Reactor Pressure Vessel / J. Schuhknecht, U. Rindelhardt, H.W. Viehrig // Проблемы прочности. — 2010. — № 1. — С. 95-104. — Бібліогр.: 9 назв. — англ. |
Institution
Digital Library of Periodicals of National Academy of Sciences of UkraineSimilar Items
Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
by: Brumovsky, M., et al.
Published: (2007)
by: Brumovsky, M., et al.
Published: (2007)
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading
by: Grinik, E.U., et al.
Published: (2003)
by: Grinik, E.U., et al.
Published: (2003)
Metalophysical problems of WWER reactor vessel welding joints reliability
by: I. H. Sharaievskyi, et al.
Published: (2022)
by: I. H. Sharaievskyi, et al.
Published: (2022)
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation
by: Dudra, Ju., et al.
Published: (2010)
by: Dudra, Ju., et al.
Published: (2010)
Validation of the code package MCPV adapted for neutron transport calculation within WWER-440 reactor near-vessel space
by: A. M. Pugach, et al.
Published: (2019)
by: A. M. Pugach, et al.
Published: (2019)
Integrated Surveillance Specimen Program for WWER-1000/V-320 Reactor Pressure Vessels
by: Brumovský, M., et al.
Published: (2007)
by: Brumovský, M., et al.
Published: (2007)
The Use of Master Curve Method for Statistical Re-Evaluation of Surveillance Test Data for WWER-1000 Reactor Pressure Vessels
by: Revka, V.M., et al.
Published: (2010)
by: Revka, V.M., et al.
Published: (2010)
Estimation of the error of functionals calculation results of the neutron flux onto the WWER 1000 reactor pressure vessel
by: V. L. Demekhin, et al.
Published: (2013)
by: V. L. Demekhin, et al.
Published: (2013)
Estimation of the error of functionals calculation results of the neutron flux onto the WWER 1000 reactor pressure vessel
by: V. L. Demekhin, et al.
Published: (2012)
by: V. L. Demekhin, et al.
Published: (2012)
Comparison of Маster curve with normative method of estimating WWER-1000 reactor pressure vessel metal fracture toughness
by: V. M. Revka, et al.
Published: (2024)
by: V. M. Revka, et al.
Published: (2024)
Mathematical modeling of residual stresses in composite welded joints of WWER-1000 reactor vessel cover with CPS nozzles
by: A. A. Makarenko, et al.
Published: (2022)
by: A. A. Makarenko, et al.
Published: (2022)
A Data Scatter for a Shift of the Ductile to Brittle Transition Temperature for WWER¬1000 Reactor Pressure Vessel Materials
by: V. N. Revka
Published: (2018)
by: V. N. Revka
Published: (2018)
Uncertainty determination of fast neutron fluence onto the WWER pressure vessel metal surveillance specimens
by: O. M. Puhach, et al.
Published: (2021)
by: O. M. Puhach, et al.
Published: (2021)
Generation of MeV -energy protons in WWER reactor core
by: Gann, A.V., et al.
Published: (2009)
by: Gann, A.V., et al.
Published: (2009)
Different approaches to estimation of reactor pressure vessel material embrittlement
by: V. M. Revka, et al.
Published: (2013)
by: V. M. Revka, et al.
Published: (2013)
Reactor Pressure Vessel and Internals Steels Irradiation Performed at the LVR-15 Research Reactor
by: Erben, Oldřich, et al.
Published: (2001)
by: Erben, Oldřich, et al.
Published: (2001)
Local approach to fracture based prediction of reactor pressure vessel lifetime
by: Kotrechko, S.A., et al.
Published: (2009)
by: Kotrechko, S.A., et al.
Published: (2009)
Impact of technological parameters of arc deposition of an anti-corrosion layer in the vessel of WWER-1000 reactor on residual stress distribution
by: O. V. Makhnenko, et al.
Published: (2020)
by: O. V. Makhnenko, et al.
Published: (2020)
SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement
by: Lucon, E., et al.
Published: (2004)
by: Lucon, E., et al.
Published: (2004)
Analysis of Warm Prestressing Effect on Fracture Toughness of Reactor Pressure Vessel Steels
by: Beleznai, R., et al.
Published: (2010)
by: Beleznai, R., et al.
Published: (2010)
Adaptation of Helios Models for WWER¬440 Fuel Assemblies for Application by the TRACE/PARCS Program
by: Yu. M. Ovdiienko, et al.
Published: (2019)
by: Yu. M. Ovdiienko, et al.
Published: (2019)
Lateral expansion and impact toughness correlation of VVER-1000 reactor pressure vessel materials
by: O. V. Tryhubenko
Published: (2014)
by: O. V. Tryhubenko
Published: (2014)
Braze-welded tubular billets for pipelines and high-pressure vessels
by: A. A. Pismennyj, et al.
Published: (2014)
by: A. A. Pismennyj, et al.
Published: (2014)
Neutron Embrittlement of WWER Reactors: EC-Supported Projects
by: Ahlstrand, R., et al.
Published: (2004)
by: Ahlstrand, R., et al.
Published: (2004)
Redistribution of residual welding stresses inside the inner-body shaft of the reactor WWER-1000 in the process of service
by: O. V. Makhnenko, et al.
Published: (2014)
by: O. V. Makhnenko, et al.
Published: (2014)
Validation of WWER-440/213 Thermohydraulic Model for TRACE Computer Code Based on RNPP-1 Incident Data
by: S. E. Yanovskyi, et al.
Published: (2019)
by: S. E. Yanovskyi, et al.
Published: (2019)
Empirical predicted residual life of the base metal of MCP of WWER-1000 reactors in operation
by: Gozhenko, S.V.
Published: (2023)
by: Gozhenko, S.V.
Published: (2023)
Analysis of criticality of melt during severe accidents in reactor vessel
by: Yu. P. Kovbasenko, et al.
Published: (2018)
by: Yu. P. Kovbasenko, et al.
Published: (2018)
Dimensioness method of assessing the conditions of thermal shock to the reactor vessel
by: V. I. Skalozubov, et al.
Published: (2014)
by: V. I. Skalozubov, et al.
Published: (2014)
Radiation embrittlement of reactor pressure vessel materials of Rivne NPP unit 1 due to re-irradiation after recovery annealing
by: M. H. Holiak, et al.
Published: (2019)
by: M. H. Holiak, et al.
Published: (2019)
Mathematical simulation of microstructure phase transformations at welding heating by example corrosion resistant cladding of the reactor vessel VVER-1000
by: O. V. Makhnenko, et al.
Published: (2018)
by: O. V. Makhnenko, et al.
Published: (2018)
Testing of pressure vessels by an international expert team
by: Ja. Nedoseka, et al.
Published: (2016)
by: Ja. Nedoseka, et al.
Published: (2016)
Effect of decarbonization of metal on the load-carrying capacity of cylindrical pressure vessels
by: Shirshov, V.P.
Published: (1985)
by: Shirshov, V.P.
Published: (1985)
Study of WWER reactors neutronic noise spectral images in irregular thermohydraulic regimes of core zones
by: I. H. Sharaievskyi, et al.
Published: (2022)
by: I. H. Sharaievskyi, et al.
Published: (2022)
Кременчуку — 440 років
Published: (2011)
Published: (2011)
Mathematical modeling of residual stresses in a composite welded joint of the collector adapter sleeve to the branch pipe of ZPM-440 steam generator
by: A. A. Makarenko, et al.
Published: (2023)
by: A. A. Makarenko, et al.
Published: (2023)
High-performance methods for analyzing the statistical strength of welded pipelines and pressure vessels using the Monte–Carlo method
by: E. A. Velikoivanenko, et al.
Published: (2020)
by: E. A. Velikoivanenko, et al.
Published: (2020)
Influence of residual process stresses on brittle fracture resistance of WWER-1000 reactor baffle in case of an emergency
by: O. V. Makhnenko, et al.
Published: (2022)
by: O. V. Makhnenko, et al.
Published: (2022)
Models of WWER-1000 nuclear reactor with division into zones on verti-cal axis for information technology of control
by: V. P. Severin, et al.
Published: (2021)
by: V. P. Severin, et al.
Published: (2021)
Radiation tests of products made of calcium-thermal zirconium grade STZ-110 under operation of the VVER-440 reactor
by: A. P. Mukhachev, et al.
Published: (2020)
by: A. P. Mukhachev, et al.
Published: (2020)
Similar Items
-
Prediction of irradiation embrittlement in WWER-440 Reactor Pressure Vessel materials
by: Brumovsky, M., et al.
Published: (2007) -
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading
by: Grinik, E.U., et al.
Published: (2003) -
Metalophysical problems of WWER reactor vessel welding joints reliability
by: I. H. Sharaievskyi, et al.
Published: (2022) -
Lifetime Analysis of WWER Reactor Pressure Vessel Internals Concerning Material Degradation
by: Dudra, Ju., et al.
Published: (2010) -
Validation of the code package MCPV adapted for neutron transport calculation within WWER-440 reactor near-vessel space
by: A. M. Pugach, et al.
Published: (2019)