Radiation hardening in magnox pressure-vessel steels

1985 ◽  
Vol 315 (1532) ◽  
pp. 301-332 ◽  
Keyword(s):  
Transition Temperature ◽  
Brittle Fracture ◽  
Yield Stress ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Spherical Particles ◽  
Irradiation Hardening ◽  
Copper Precipitation ◽  
Pressure Vessel Steels

The ferritic steels used for reactor pressure vessels undergo a marked transition from ductile to brittle fracture behaviour over a relatively narrow temperature range. For most unirradiated mild steels the ductile to brittle transition temperature (d.b.t.t.) is between — 50° and 20 °C. The process of irradiation hardening, through the formation of clusters of interstitial or vacancy defects, increases the friction stress of these steels and thereby raises the transition temperature. Given the inherent tendency of these steels to fail in a brittle manner, the raising of the transition temperature under neutron irradiation poses a problem of considerable technological importance in the nuclear industry. At the time (1962) when the first of the Central Electricity Generating Board (C.E.G.B.) Magnox nuclear stations began operation the phenomenon of brittle fracture was already comparatively well understood. A theory of the process had already been developed and applied to the problem of radiation embrittlement. However, as the results from the Magnox pressure-vessel surveillance scheme accumulated, it gradually became evident that the measured changes in yield stress in the monitoring specimens could not be accounted for simply on the basis of irradiation hardening through the formation of damage clusters. By the late 1970s, sufficient data had been gathered from the surveillance programme to enable a detailed investigation of the processes occurring in the Magnox steels to be instituted. The form of the investigation was to subsequently evolve into two phases; an initial com prehensive microstructural study of the steels, followed by the formation of an interpretative model based on the observations. In this paper we present the Magnox yield-stress monitoring measurements and then briefly describe the principal findings from our microstructural studies. The Magnox pressure-vessel steels contain between 0.05 and 0 .4 % by mass of copper and we show that under certain conditions this element may precipitate as small spherical particles within the matrix of the steels. A review of previous work on copper precipitation in ferrite is then followed by a description of our model. This assumes that the changes in yield stress generally arise from the combined effects of irradiation damage loops and copper precipitates. The formation of the latter may be enhanced by irradiation and in some steels their contribution is dominant. It is shown that the model successfully accounts for the measurements made on both plate and weld steels in all the Magnox stations. Experimental support for the model comes from our own microstructural observations and from other studies, in the U.K. and elsewhere, using techniques which allow the detection of sub-microscopic particles in steels. The model may be applied to pressure-vessel steels in other reactor systems. Indeed, it predicts that the yield-stress changes in steels with a high copper content irradiated under p.w.r. (pressurized water reactor) conditions will be dominated by the contribution from copper precipitation.

Modelling of Irradiation Embrittlement in the Ductile to Brittle Transition Range for an A508 Pressure Vessel Steel

2006 ◽  
Author(s):  
B. Tanguy ◽  
J. Besson ◽  
S. Bugat
Keyword(s):  
Brittle Fracture ◽  
Ductile Fracture ◽  
Yield Stress ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Transition Curve ◽  
Dissipated Energy ◽  
Pressure Vessel Steel ◽  
Transition Range ◽  
Vessel Steel

The aging behavior of structural steels used to manufacture nuclear pressure vessels is surveyed using Charpy V-notch specimens located in capsules inside the pressure vessel. The Charpy data are then used to assess the safety integrity of the structures based on semi-empirical relations relating Charpy impact transition curve shifts and the fracture toughness shifts due to irradiation. Using a computational strategy proposed in [1] which combines a deterministic model for ductile fracture and a statistical description of brittle fracture, this work aims at the prediction of the whole Charpy transition curve of irradiated steels. The actual strain hardening behavior of an A508 Cl.3 steel from the french surveillance program is considered in the simulations, contrarily to a previous work where a shift of the un-irradiated stress-strain curve to higher stress values was considered. Comparison with Charpy energy data for two levels of irradiation shows that irradiation possibly also affect brittle fracture. It is also shown that if a low increase of the yield stress is considered, the ductile fracture energy can decrease as a result of a compensation between the increase of dissipated energy due to a higher yield stress and a decrease of dissipated energy due to a faster ductile crack propagation.

Experimental Program to Monitor the Irradiation Induced Embrittlement of the French Reactor Pressure Vessel Steels

2004 ◽  
Author(s):  
Guillaume Chas ◽  
Nathalie Rupa ◽  
Josseline Bourgoin ◽  
Astrid Hotellier ◽  
Se´bastien Saillet
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Mechanical Tests ◽  
Nuclear Industry ◽  
Experimental Program ◽  
Surveillance Program ◽  
Base Metals ◽  
Vessel Steel ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

By monitoring the irradiation-induced embrittlement of materials, the Pressure Vessel Surveillance Program (PVSP) contributes to the RPV integrity and lifetime assessments. This program is implemented on each PWR Unit in France; it is mainly based on Charpy tests, which are widely used in the nuclear industry to characterize the mechanical properties of the materials. Moreover, toughness tests are also carried out to check the conservatism of the PVSP methodology. This paper first describes the procedure followed for the Pressure Vessel Surveillance Program. It presents the irradiation capsules: the samples materials (low alloy Mn, Ni, Mo vessel steel including base metals, heat affected zones, welds and a reference material) and the mechanical tests performed. Then it draws up a synthesis of the analysis of about 180 capsules removed from the reactors at fluence levels up to 7.1019 n/cm2 (E > 1 MeV). This database gathers the results of more than 10,000 Charpy tests and 250 toughness tests. The experimental results confirm the conservatism of the Code-based methodology applied to the toughness assessment.

Fracture Behavior in West Jefferson Test Under Low-Temperature Condition for X65 Steel Pipe With High Charpy Energy: Current Activities in HLP Committee, Japan, Report 1

2016 ◽  
Author(s):  
Toshihiko Amano ◽  
Satoshi Igi ◽  
Takahiro Sakimoto ◽  
Takehiro Inoue ◽  
Shuji Aihara
Keyword(s):  
Transition Temperature ◽  
Low Temperature ◽  
Brittle Fracture ◽  
Pressure Vessel ◽  
Fracture Behavior ◽  
Line Pipe ◽  
Burst Test ◽  
Pipe Burst ◽  
Test Vessel ◽  
Fracture Appearance

This paper describes the results of pressure vessel fracture test which called West Jefferson and/or partial gas burst testing using Grade API X65 linepipe steel with high Charpy energy that exhibits inverse facture in the Drop Weight Tear Test (DWTT). A series of pressure vessel fracture tests which is as part of an ongoing effort by the High-strength Line Pipe committee (HLP) of the Iron and Steel Institute of Japan (ISIJ) was carried out at low temperature in order to investigate brittle-to-ductile transition behavior and to compare to DWTT fracture behavior. Two different materials on Fracture Appearance Transition Temperature (FATT) property were used in these tests. One is −60 degree C and the other is −25 to −30 degree C which is defined as 85 % shear area fraction (SA) in the standard pressed notch DWTT (PN-DWTT). The dimensions of the test pipes were 24inches (609.6 mm) in outside diameter (OD), 19.1 mm in wall thickness (WT). In each test, the test pipe is cooled by using liquid nitrogen in the cooling baths. Two cooling baths are set up separately on the two sides of the test vessel, making it possible to obtain fracture behaviors under two different test temperatures in one burst test. The test vessel was also instrumented with pressure transducers, thermocouples and timing wires to obtain the pressure at the fracture onset, temperature and crack propagation velocity, respectively. Some informative observations to discuss appropriate evaluation method for material resistance to brittle facture propagation for high toughness linepipe materials are obtained in the test. When the pipe burst test temperatures are higher than the PN-DWTT transition temperature, ductile cracks were initiated from the initial notch and propagated with short distance in ductile manner. When the pipe burst test temperatures were lower than the PN-DWTT transition temperature, brittle cracks were initiated from the initial notch and propagated through cooling bath. However, the initiated ductile crack at lower than the transition temperature was not changed to brittle manner. This means inverse facture occurred in the PN-DWTT is a particular problem caused by the API DWTT testing method. Furthermore, results for the pipes tested indicated that inverse facture occurred in PN-DWTT at the temperature above the 85 % FATT may not affect the arrestability against the brittle fracture propagation and it is closely related with the location of brittle fracture initiation origin in the fracture appearance of PN-DWTT.

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.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Brittle Fracture Assessments on Piping Systems: MSOT Curves

2019 ◽  
Author(s):  
Ishita Chakraborty ◽  
Kannan Subramanian ◽  
Jorge Penso
Keyword(s):  
Brittle Fracture ◽  
Case Studies ◽  
Pressure Vessel ◽  
Iterative Process ◽  
Pressure Vessels ◽  
Temperature Changes ◽  
End Conditions ◽  
Piping Systems ◽  
Plant Personnel ◽  
Safe Operating

Abstract Brittle fracture assessments (BFAs) of pressure vessels based on API 579-1/ASME FFS-1, Section 3 procedures are frequently easier and more straightforward to implement in comparison to the BFAs on piping systems. Specifically, the development of the MSOT curves. This is due to the complexities involved in the piping systems due to the branch piping interactions, end conditions of piping systems such as nozzle flexibilities at the pressure vessel connections, temperature changes in the length of piping especially when the piping is significantly long as seen in flare header piping systems. MSOT curves that are alternatively used for MAT curves provide a better picture to the plant personnel in understanding the safe operating envelope. Development of MSOT curves is an iterative process and therefore involves significant number of piping stress analyses during their development. In this paper, an approach to develop the MSOT curves is discussed with two case studies that are of relevance to olefin plants.

Conservative estimations of irradiation embrittlement of reactor pressure vessel steels for WWER-1000 lifetime prediction

2019 ◽  
Vol 15 (1) ◽  
pp. 246-257
Author(s):  
Nikolai Petrovich Anosov ◽  
Vladimir Nikolaevich Skorobogatykh ◽  
Lyubov’ Yur’yevna Gordyuk ◽  
Vasilii Anatol’evich Mikheev ◽  
Egor Vasil’yevich Pogorelov ◽  
...  
Keyword(s):  
Transition Temperature ◽  
Brittle Fracture ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Lifetime Prediction ◽  
Critical Brittleness Temperature ◽  
Irradiation Embrittlement ◽  
Content Type ◽  
Brittleness Temperature ◽  
Reactor Pressure

Purpose The purpose of this paper is to consider a procedure of water-water energetic reactor (WWER) reactor pressure vessel (RPV) lifetime prediction at the stages of design and lifetime extension using the standard irradiation embrittlement parameters as defined in regulatory documents. A comparison is made of the brittle fracture resistance (BFR) values evaluated using two criteria: shift in the critical brittleness temperature ΔTc or shift in the brittle-to-ductile transition temperature ΔTp and without shifts (Tc and Tp). Design/methodology/approach The radiation resistance was determined using the following three approaches: calculation based on standard values ΔTc and Tc0 or ΔTp and Tp0 (a level of excessive conservatism); calculation based on standard value ΔTc and actual value Tc0 or actual values ΔTp and Tp0 (the level of realistic conservatism); or calculation based on actual values of Tc and Tc0 or Tp and Tp0 (the level of actual conservatism). The BFR was evaluated based on the results of testing the specimens subjected to irradiation in research reactors as well as surveillance specimens subjected to irradiation immediately under operating conditions. Findings The excessive conservatism in determining the actual lifetime of nuclear reactor vessel materials can be eliminated by using the immediate values of critical brittleness temperature and ductile-to-brittle transition temperature. Originality/value Obtained results can be applied to extend WWER vessel operating time at the stages of designing and operation due to substantiated decrease in conservatism. And it will allow carrying out a statistical substantiated assessment of the resistance to brittle fracture of the RPV steels.

The Assessment and Assurance of Pressure Vessel Reliability

1975 ◽  
Vol 189 (1) ◽  
pp. 391-404 ◽  
Author(s):  
R. W. Nichols
Keyword(s):  
Quality Assurance ◽  
Pressure Vessel ◽  
Reliability Assessment ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Detailed Design ◽  
Confidence Levels ◽  
Statistical Confidence ◽  
Reliability Assessments ◽  
Operational Behaviour

The factors involved in assessing the reliability of pressure vessels drawing extensively upon the developments which have arisen from applications in the nuclear industry. Existing assessments of reliability and operational behaviour highlight some improvements which could result from more detailed design assessments especially with respect to stress analysis, stress transients and the significance of defects. Additionally the contributions to reliability made by fabrication and materials technology, inspection and quality assurance and post operational surveillance are critically examined. The use of such data in synthesizing a reliability assessment is discussed noting the problems of establishing statistical confidence levels and highlighting those areas where further evidence would produce significant advances in quantifying reliability assessments.

Prediction of Fracture Toughness KIC Transition Curves of Pressure Vessel Steels From Charpy V-Notch Impact Test Results

1994 ◽  
Vol 116 (4) ◽  
pp. 353-358 ◽  
Author(s):  
T. Iwadate ◽  
Y. Tanaka ◽  
H. Takemata
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Prediction Method ◽  
Pressure Vessels ◽  
Life Extension ◽  
Low Alloy Steels ◽  
Excess Temperature ◽  
Pressure Vessel Steels ◽  
Transition Curves

A single and generalized prediction method of fracture toughness KIC transition curves of pressure vessel steels has been greatly desired by engineers in the petro-chemical and nuclear power industries, especially from the viewpoint of life extension of reactor pressure vessels. In this paper, the toughness degradation of Cr-Mo steels during long-term service was examined and the two prediction methods of fracture toughness KIC transition curves were studied using the data of 54 heats. 1) The toughness degradation of 2 1/4Cr-1Mo steels levels off within around 50,000 h service. 2) The FATT versus J-factor (=(Si+Mn)(P+Sn)×104) and/or X (=(10P+5Sb+4Sn+As)x10−2) relationships to estimate the maximum embrittlement of Cr-Mo steels were obtained. 3) A master curve method developed by authors et al.; that is, the method using a KIC/KIC−US versus excess temperature master curve of each material was presented for 2 1/4Cr-1Mo, 1 1/4Cr-1/2Mo, 1Cr and 1/2Mo chemical pressure vessel steels and ASTM A508 C1.1, A508 C1.2, A508 C1.3 and A533 Gr.B C1.1 nuclear pressure vessel steels, where KIC−US is the upper-shelf fracture toughness and excess temperature is test temperature minus FATT. 4) A generalized prediction method to predict the KIC transition curves of any low-alloy steels was developed. This method consists of KIC/KIC−US versus T–T0 master curve and temperature shift ΔT between fracture toughness and CVN impact transition curves versus yield strength relationship, where To is the temperature showing 50 percent KIC−US of the material. 5) The KIC transition curves predicted using both methods showed a good agreement with the lower bound of measured KJC values obtained from JC tests.

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