To Determination of the WWER RPV Steels Crack Resistance Characteristics

2013 ◽  
Author(s):  
Igor Orynyak ◽  
Maksym Zarazovskii ◽  
Sergii Radchenko ◽  
Volodymyr Kozlov
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Crack Resistance ◽  
Temperature Shift ◽  
Charpy Impact ◽  
Temperature Range ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure ◽  
15Kh2nmfa Steel

The efficiency of fracture toughness determined by the methodology of normative document PNAE G-7-002-86 has been analyzed. Crack resistance characteristics of WWER-1000 reactor pressure vessel base metal at unirradiated condition are obtained by experimental way. All specimens were made of the RPV support forging (15Kh2NMFA steel) of abandoned Crimean NPP Fracture toughness experiments were carried out on three types of specimens CT 1T, CT 0.5T and SEB over a temperature range from −130°C to −40°C in fully accordance to the ASTM E1921. Charpy impact energy data obtained on twelve specimens over a temperature range from −80°C to 80°C has been used to determine the 47J transition temperature. Comparison of obtained fracture toughness data with normative curve shows that the last one has unreasonably high lower shelf. It has been found that the PNAE G-7-002-86 Code, which uses the ideology of transition temperature shift, is too conservative to estimate WWER-1000 RPVs resistance against brittle fracture for the pressurized thermal shock (over 90 MPa·√m area of stress intensity factor).

A Proposal for the Maximum KIc for Use in ASME Code Flaw and Fracture Toughness Evaluations

2011 ◽  
Author(s):  
Mark T. Kirk ◽  
Gary L. Stevens ◽  
Marjorie Erickson ◽  
Shengjun Sean Yin
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Empirical Models ◽  
Ferritic Steels ◽  
Transition Regime ◽  
Upper Limit ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
The One ◽  
Reactor Pressure

Nonmandatory Appendices A [1] and G [2] of Section XI of the ASME Code use the KIc curve (indexed to the material reference transition temperature, RTNDT) in reactor pressure vessel (RPV) flaw evaluations, and for the purpose of establishing RPV pressure-temperature (P-T) limits. Neither of these appendices places an upper-limit on the KIc value that may be used in these assessments. Over the years, it has often been suggested by some of the members of the ASME Section XI Code committees that are responsible for maintaining Appendices A and G that there is a practical upper limit of 200 ksi√in (220 MPa√m) [4]. This upper limit is not well recognized by all users of the ASME Code, is not explicitly documented within the Code itself, and the one source known to the authors where it is defended [4] relies on data that is either in error, or is less than 220 MPa√m. However, as part of the NRC/industry pressurized thermal shock (PTS) reevaluation effort, empirical models were developed that propose common temperature dependencies for all ferritic steels operating on the upper shelf. These models relate the fracture toughness properties in the transition regime to those on the upper shelf and, combined with data for a wide variety of RPV steels and welds on which they are based, suggest that the practical upper limit of 220 MPa√m exceeds the upper shelf fracture toughness of most RPV steels by a considerable amount, especially for irradiated steels. In this paper, available models and data are used to propose upper bound limits of applicability on the KIc curve for use in ASME Code, Section XI, Nonmandatory Appendices A and G evaluations that are consistent with available data for RPV steels.

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.

Neutron Embrittlement Aging Management of Nuclear Reactor Pressure Vessels

2003 ◽  
Author(s):  
William L. Server
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Temperature Shift ◽  
Environmental Parameters ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Vessel Material ◽  
Reactor Pressure

The management of neutron embrittlement of nuclear reactor pressure vessels involves monitoring of the changes in the fracture toughness of surveillance capsule specimens that closely approximate the actual reactor vessel material(s). The measurement of fracture toughness is currently performed in an indirect manner using Charpy V-notch impact specimens, although the direct measurement of fracture toughness is possible using the same small Charpy specimens fatigue precracked to produce acceptable fracture toughness three-point bend specimens. This paper first examines the current Charpy-based approach and the development of a recent embrittlement correlation that has been incorporated into ASTM E 900-02, “Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials.” This correlation provides the latest mechanistically-guided approach to assess the changes in transition temperature shift. This same correlation and mechanistic guidance can be used with measured fracture toughness data developed following ASTM E 1921-02 to account for differences in surveillance material versus actual vessel material. Additionally, environmental parameters such as fluence and temperature also can be adjusted between different irradiation facilities using this latest correlation. This paper focuses on the application of the new ASTM E 900-02 correlation to Charpy-based and fracture toughness-based measurements to develop the best predictive approach for assuring structural integrity of reactor vessel materials. Key technical issues important for extended vessel life also are discussed.

Fracture Toughness of Highly Irradiated Pressure Vessel Steels in the Upper Shelf Temperature

2006 ◽  
Author(s):  
Hiroshi Matsuzawa ◽  
Toru Osaki
Keyword(s):  
Fracture Toughness ◽  
Chemical Composition ◽  
Reference Material ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Shelf Energy ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Nine Reactor Pressure Vessel (RPV) Steels and four RPV weld were irradiated up to 1.2 × 1024n/m2 fast neutron fluence (E>1MeV), and their fracture toughness and Charpy impact energy were measured. As chemical compositions, such as Cu, are known to affect the fracture toughness reduction due to neutron exposure, the above steels were fabricated by changing chemical composition widely to cover the chemical composition of the RPV materials of the operating Japanese nuclear power plants. 2.7 mm thick compact specimens were used to measure the upper shelf fracture toughness of highly irradiated materials, and their Charpy upper shelf energy was also measured. By correlating Charpy upper shelf energy to fracture toughness, the upper shelf fracture toughness evaluation formulae for highly irradiated reactor pressure vessel steels were developed. Both compact and V-notched Charpy impact specimens were irradiated in a test reactor. The fast neutron flux above 1MeV was about 5 × 1016n/(m2s). Charpy impact specimens made of Japanese PWR reference material containing 0.09w% Cu were irradiated simultaneously. The upper shelf energy of the reference material up to the medium fluence level showed little difference in the reduction of upper shelf energy to that which had been in the operating plant and which was irradiated to the same fluence. The developed correlation formulae have been adopted in the Japan Electric Association Code as new formulae to predict the fracture toughness in the upper shelf region of reactor pressure vessels. They will be applied to time limited ageing analysis of low upper shelf reactor pressure vessels in Japan, on a concrete technical basis in very high fluence regions.

Evaluation of Fracture Toughness From Instrumented Charpy Impact Tests for a Reactor Pressure Vessel Steel

2003 ◽  
Author(s):  
A. Parrot ◽  
P. Forget ◽  
A. Dahl
Keyword(s):  
Fracture Toughness ◽  
Impact Test ◽  
Charpy Impact ◽  
Charpy Impact Test ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Impact Tests ◽  
Reactor Pressure Vessel Steel ◽  
Ductile To Brittle Transition ◽  
Reactor Pressure

The monitoring of neutron induced embrittlement of nuclear power plants is provided using Charpy impact test in the surveillance program. However structural integrity assessments require the fracture toughness. Some empirical formulas have been developed but no direct relationship was found. The aim of our study is to determine the fracture toughness of a Reactor Pressure Vessel steel from instrumented Charpy impact test using local approach to fracture. This non-empirical method has been applied in the brittle domain as well as in the ductile to brittle transition for an A508 C1.3 steel. In the brittle domain, fracture occurs by cleavage and can be modeled with the Beremin model. Fracture toughness has been successfully determined from Charpy impact tests results and the influence of several parameters (mesh design, Beremin model with one or two parameters, number of Charpy impact tests results) on the results was considered. In the ductile to brittle transition, cleavage fracture is preceded by ductile crack growth. Ductile tearing has been accounted for in the simulations with the Rousselier model whereas cleavage fracture is still described with the Beremin model. The determination of fracture toughness from Charpy impact tests gave encouraging results but finite element simulations have to be refined in order to improve predictions.

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