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JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in pressurized water reactors (Contract research)

Kasahara, Shigeki; Fukuya, Koji*; Fujimoto, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-013, 171 Pages, 2019/01

JAEA-Review-2018-013.pdf:6.89MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. When the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of pressurized water reactors (PWR) and boiling water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of PWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of PWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into tables.

JAEA Reports

Data survey and compilation of material property tables of irradiated stainless steel for evaluation of radiation effects on structural material properties of core internals in boiling water reactors (Contract Research)

Kasahara, Shigeki; Fukuya, Koji*; Koshiishi, Masato*; Fujii, Katsuhiko*; Chimi, Yasuhiro

JAEA-Review 2018-012, 180 Pages, 2018/11

JAEA-Review-2018-012.pdf:10.71MB

For structural integrity assessment of reactor internals of light water reactors, it is important to evaluate and predict the property changes of structural materials, based on existing data obtained from austenitic stainless steel irradiated with neutrons. Compilation of the data into tables is valuable for discussing the representative or the most probable values of the properties applied to the assessment. In the process of the data compilation, the data must be distinguished clearly in consideration of different service conditions of core internals of boiling water reactors (BWR) and pressurized water reactors. Main objective of this work is to provide material property tables of irradiated austenitic stainless steel which will be applicable for assessment of structural integrity of core internals of BWRs. To compile the table, published literature reporting irradiated stainless steel data were surveyed and screened by considering the service conditions of BWRs. In addition to the data, various parameters for the data evaluation, e.g. chemical compositions and pre-treatments of the materials, irradiation and examination conditions, were extracted from the literature, and listed into the tables.

Oral presentation

Data survey and trend analysis of radiation effects on structural material properties of core internals in light water reactors, 3; IASCC propagation and fracture toughness

Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Koshiishi, Masato*

no journal, , 

no abstracts in English

Oral presentation

Data survey and trend analysis of radiation effects on structural material properties of core internals in light water reactors, 2; Tensile properties and IASCC initiation

Fukuya, Koji*; Chimi, Yasuhiro; Kasahara, Shigeki; Fujii, Katsuhiko*; Fujimoto, Koji*

no journal, , 

no abstracts in English

Oral presentation

Data survey and trend analysis of radiation effects on structural material properties of core internals in light water reactors, 1; Overview

Chimi, Yasuhiro; Fukuya, Koji*; Kasahara, Shigeki; Fujii, Katsuhiko*; Hanawa, Satoshi

no journal, , 

To contribute to materials degradation assessment in core internals of light water reactors, we have widely surveyed the literature on irradiation properties of austenitic stainless steels (SSs) and summarized the data. In the present investigation, in terms of mechanical properties (tensile properties, hardness, and fracture toughness), IASCC properties (IASCC susceptibility, IASCC initiation, and IASCC growth), stress relaxation - creep - swelling, and microstructural properties (microstructures and grain boundary segregation) in irradiated SSs, we have collected the data and made the spread sheets separately on PWR and BWR according to the test conditions and the target materials. We have also considered the trend curves on the mechanical properties, the IASCC properties, etc. based on the knowledge obtained through surveying the literature on dose dependence of the irradiation properties. In the presentation, we will report the overview of the present investigation.

Oral presentation

Trend equations for dose dependence of crack growth rates of neutron irradiated austenitic stainless steel under high temperature aqueous conditions

Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Koshiishi, Masato*

no journal, , 

no abstracts in English

Oral presentation

Change in mechanical properties of austenitic stainless steels irradiated in light water reactors

Fukuya, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro

no journal, , 

To evaluate the effects of neutron irradiation on the tensile properties of austenitic stainless steels for reactor core internals in light water reactors, we have revised the database on tensile properties of irradiated stainless steels and investigated the dose dependence. Based on the fundamental equation, which can express the tendency toward saturation of tensile properties with increasing the dose, we have proposed the appropriate trend equations of dose dependence for the classified data in terms of irradiation conditions, materials, work and thermal treatment conditions, etc. In the presentation, we will report the results of investigation on the effects of each condition on the irradiation behavior of tensile properties.

Oral presentation

Empirical equations of crack growth rates of neutron irradiated stainless steel under simulated BWR core conditions

Kasahara, Shigeki; Fukuya, Koji*; Chimi, Yasuhiro; Fujii, Katsuhiko*; Hata, Kuniki

no journal, , 

Disposition curves of crack growth rates (CGR) of stainless steel in appropriate consideration of IASCC are necessary for structural integrity assessment of reactor internals of BWRs. This paper describes empirical equations development of CGR (da/dt), as functions of stress intensity factors (K) and neutron dose (F) to contribute to improvement of the structural integrity assessment. Development started from a formula of da/dt=M$$times$$K$$^{n}$$, and on the assumption that "M" and "n" tend to be saturated with increasing F. Datasets for fitting were prepared consisting of CGR, F and K from the results of PIE under simulated NWC conditions. Data fitting with least square method was applied to the datasets to obtain the equation. The results from the empirical equation were compared with the measured crack growth data, and validity of the equations were discussed from the viewpoints of statistical analysis.

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