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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 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.

Journal Articles

In-situ measurement of electrical conductivity of solution within crevice of stainless steel in high temperature and high purity water

Soma, Yasutaka; Komatsu, Atsushi; Ueno, Fumiyoshi

Zairyo To Kankyo, 67(9), p.381 - 385, 2018/09

In-situ measurement of electrical conductivity of solution within crevice of SUS316L stainless steel in 288$$^{circ}$$C water has been conducted with newly developed electrochemical sensor system. The sensor measures local electrical conductivity of crevice solution beneath the electrode ($$kappa$$$$_{crev}$$) with electrochemical impedance method. The sensors were installed at different positions within tapered crevice of SUS316L stainless steel. The crevice specimen with the sensors were immerged into 288$$^{circ}$$C, 8 MPa, pure oxygen saturated high purity water for 100 h. $$kappa$$$$_{crev}$$ at a position with crevice gap of $$approx$$59.3$$mu$$m was 8-11$$mu$$S/cm, least deviate from conductivity of 288$$^{circ}$$C pure water (4.4$$mu$$S/cm) and no localized corrosion occurred. On the contrary, $$kappa$$$$_{crev}$$ at a position with crevice gap of $$approx$$4.4$$mu$$m increased with time and showed maximum value of $$approx$$1600$$mu$$S/cm at 70 h. Localized corrosion occurred in the vicinity of this position. Thermodynamic equilibrium calculation showed $$kappa$$$$_{crev}$$ of 1600$$mu$$S/cm being equivalent to pH of 3 to 3.7. It can be concluded that acidification occurred in tight crevice even under high purity bulk water and resulted in localized corrosion.

Journal Articles

Creep damage evaluations for BWR lower head in severe accident

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko

Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08

It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.

JAEA Reports

None

Ohara, Hiroshi*; *

PNC-TJ8164 96-009, 261 Pages, 1996/09

PNC-TJ8164-96-009.pdf:13.68MB

no abstracts in English

JAEA Reports

Oral presentation

Development of numerical simulation for jet break up behavior in complicated structure of BWR lower plenum, 12; Evaluation of jet breakup length in complicated structures using LIF method

Narushima, Yuki*; Abe, Yutaka*; Kaneko, Akiko*; Kanagawa, Tetsuya*; Suzuki, Takayuki*; Yoshida, Hiroyuki

no journal, , 

no abstracts in English

Oral presentation

Consideration of material analysis using simulated fuel assembly heating test, 1; Outline of evaluation in simulated fuel assembly heating test

Abe, Yuta; Nakagiri, Toshio; Sato, Ikken; Nakano, Natsuko*; Yamaguchi, Hidenobu*

no journal, , 

no abstracts in English

Oral presentation

Analysis of core temperature increase behavior in Fukushima Daiichi NPP Unit 2 by RELAP/SCDAPSIM

Yoshikawa, Shinji; Sato, Ikken

no journal, , 

In the accidents of the Fukushima Dai-ichi Nuclear Power Plant (FDNPP), in contrast that core degradations are thought to begin before depressurization in Unit 1 and 3, the core of Unit 2 was presumed to be intact at the time of depressurization and the degradation is thought to begin when the liquid level was close to or below the bottom of the core. The authors analyzed a temperature increase behavior during the core degradation of Unit 2 based on simulations with RELAP/SCDAPSIM computer code.

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