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Journal Articles

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.

Journal of Nuclear Materials, 487, p.229 - 237, 2017/04

 Times Cited Count:27 Percentile:97.79(Materials Science, Multidisciplinary)

Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400$$^{circ}$$C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200$$^{circ}$$C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200$$^{circ}$$C. This degradation was attributed to grain boundary sliding deformation with $$gamma$$/$$delta$$ transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 $$^{circ}$$C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

Journal Articles

Evaluation on tolerance to failure of ODS ferritic steel claddings at the accident conditions of fast reactors

Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.

JAEA Reports

Development of the Joyo MK-II core bowing reactivity calculation code

; ; Aoyama, Takafumi; Torimaru, Tadahiko

JNC TN9410 99-018, 34 Pages, 1999/09

JNC-TN9410-99-018.pdf:1.49MB

The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity has been calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occured at isolated core component, such as control rod or irradiation rig and at the interface region of fuel and reflector where had sharp material reactivity worth gradient.

Journal Articles

None

Aoyama, Takafumi; Torimaru, Tadahiko; ; Arii, Yoshio;

Nihon Genshiryoku Gakkai-Shi, 41(9), p.946 - 953, 1999/00

 Times Cited Count:1 Percentile:13.53(Nuclear Science & Technology)

None

JAEA Reports

Measurement and evaluation of decay heat on the "JOYO" spent fuel; Decay heat of short term cooled spent fuel

Torimaru, Tadahiko; Yoshida, Akihiro; Nagasaki, Hideaki*; Suzuki, Soju

PNC TN9410 98-034, 31 Pages, 1998/03

PNC-TN9410-98-034.pdf:0.58MB

Decay heat measurement system for the JOYO spent fuels was developed to obtain the decay heat data of the fuel assemblies as a non-destructive examination method. Since then, decay heat of the JOYO Mk-II fuels, which were cooled for more than 70 days, was measured in the spent fuel storage pond. The measurement of the short term cooled spent fuels, which were discharged without cooling in the reactor vessel, was performed in order to obtain the higher decay heat of the spent fuels. The burn-up of the measured fuels was about 60GWd/t, and the shortest cooling time was 24 days. The experimental data were compared with calculated values of ORIGEN2 using new libraries based on latest nuclear data library "JENDL-3.2". The main results are as follows; (1) The measured decay heat at 24 days after the reactor shut down was approximately 1.25$$pm$$0.3 kW. (2) The ratio between calculated and experimental values, C/E, was approximately 0.9 and indicated a cool down time dependence. (3) The heat generated by $$^{238}$$Pu and $$^{241}$$Am, which amount to 1% of initial composition of fresh fuel, reached 7 - 19% of decay heat at 24-160 days after the reactor shut down.

Journal Articles

None

Aoyama, Takafumi; Torimaru, Tadahiko; Nose, Shoichi; ;

Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7), , 

None

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 2; Hot-extruded bar and tube manufacturing test

Oka, Hiroshi; Tanno, Takashi; Inoue, Toshihiko; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*

no journal, , 

no abstracts in English

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 4; Evaluation of failure limit correlation under an accident condition

Yano, Yasuhide; Inoue, Toshihiko; Otsuka, Satoshi; Furukawa, Tomohiro; Kato, Shoichi; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*

no journal, , 

no abstracts in English

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 3; Mechanical properties at elevated temperature

Kato, Shoichi; Furukawa, Tomohiro; Otsuka, Satoshi; Yano, Yasuhide; Inoue, Toshihiko; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*

no journal, , 

In order to evaluate the fracture limit of the cladding material made by ODS at the severe accident condition, the mechanical strength tests have been performed at elevated temperature. In this meeting, the research plan and the progress on the mechanical strength under this research project is presented. In addition, the technical development result concerning the 1000$$^{circ}$$C creep apparatus prepared for this research is also reported.

Oral presentation

R&D of fuel cladding of ODS ferritic steel for maintaining fuel integrity at accidental high temperature condition, 2-1; Evaluation of failure limit correlation under an accident condition

Yano, Yasuhide; Kato, Shoichi; Otsuka, Satoshi; Inoue, Toshihiko; Tanno, Takashi; Oka, Hiroshi; Furukawa, Tomohiro; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; et al.

no journal, , 

no abstracts in English

Oral presentation

The Oxide particle growth of 9Cr-ODS steel claddings during ultra-high temperature tensile testing

Sowa, Takashi*; Ukai, Shigeharu*; Ono, Naoko*; Yano, Yasuhide; Kaito, Takeji; Torimaru, Tadahiko*; Kimura, Akihiko*; Hayashi, Shigenari*

no journal, , 

no abstracts in English

Oral presentation

High temperature creep properties of ODS steel cladding for evaluating severe accident

Kato, Shoichi; Furukawa, Tomohiro; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Oka, Hiroshi; Inoue, Toshihiko; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

no journal, , 

Oxide dispersion strengthened (ODS) steel is a prime candidate for cladding tubes of Japan Sodium-cooled Fast Reactor (JSFR) due to the high temperature and radiation resistances. One of the safety design of JSFR for Design Extension Condition (DEC) is the control of severe plant conditions, including prevention of severe accidents and mitigation of severe-accident consequences. Therefore, it is necessary to acquire the mechanical properties at ultra-high temperature conditions for core materials to evaluate safety design. There are, however, no data for ODS claddings at ultra-high temperature condition for the reflecting to the design criteria. In this study, creep rupture tests of 9Cr-ODS, 12Cr-ODS and FeCrAl-ODS steel claddings have been done at elevated temperatures, and the effect of minor elements such as Al, Zr and O on the mechanical strength and the creep rupture curve for the safety design were evaluated. The effect of minor elements was estimated based on the data at 700$$^{circ}$$C and 1000$$^{circ}$$C. As the results, it was confirmed that the addition of Zr had an effect on the improvement of creep strength at elevated temperature for the FeCrAl-ODS steel claddings.

Oral presentation

Transient burst properties of ODS steel cladding for evaluating sever accident

Inoue, Toshihiko; Sekio, Yoshihiro; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Furukawa, Tomohiro; Kaito, Takeji; Torimaru, Tadahiko*; Hayashi, Shigenari*; et al.

no journal, , 

In order to evaluate the strength and deformation in severe accident, the transient burst tests were carried out with various heating rates (from 0.1 to 10 K/s) and hoop stresses (from 50 to 200 MPa) to provide more evaluation data. The test materials were core materials in fast reactors, 9-18Cr-ODS and accident tolerant fuel cladding tube in the light water reactors, FeCrAl-added ODS ferritic steels. Result, the rupture strength dropped with increasing hoop stress and decreasing heating rate. The burst strength of Al-added ODS steels was lower than other ODS steels, Al and Zr-added ODS steels show good transient burst strength.

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 3-5; Evaluation on tolerance to failure of existing ODS ferritic steel claddings at the accident conditions of fast reactors

Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

no journal, , 

no abstracts in English

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 3-3; Formulation of failure life evaluation for FeCr- and FeCrAl-ODS steel claddings

Yano, Yasuhide; Kato, Shoichi; Otsuka, Satoshi; Uwaba, Tomoyuki; Sekio, Yoshihiro; Inoue, Toshihiko; Furukawa, Tomohiro; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

no journal, , 

no abstracts in English

Oral presentation

R&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 3-2; Mechanical properties of FeCr- and FeCrAl-ODS steels at elevated temperature

Kato, Shoichi; Furukawa, Tomohiro; Otsuka, Satoshi; Yano, Yasuhide; Inoue, Toshihiko; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*

no journal, , 

An evaluation on tolerance to failure of existing ODS ferritic steel claddings at the accident condition is important from the viewpoint of the reactor safety. This paper describes the high temperature strength of the 9/12Cr-ODS steels for fast reactors and the FeCrAl-ODS steels for light water reactors.

Oral presentation

Development of quick and remote analysis for severe accident reactor, 6-2; Analysis of LIBS spectra using the least-square method for simulated fuel debris

Akaoka, Katsuaki; Oba, Hironori; Wakaida, Ikuo; Ouchi, Atsushi*; Mizusako, Fumiki*; Eto, Yoshinori*; Torimaru, Tadahiko*

no journal, , 

For measurement of the fuel debris from the Fukushima Daiichi Nuclear Power Station accident, we analyzed the LIBS spectrum of the mixed simulated fuel debris of U, Zr and Fe, using the least-squares method. We present the results of the calibration curve and analysis spectrum.

Oral presentation

Development of quick and remote analysis for severe accident reactor, 6-1; Analysis of simulated fuel debris by fiber-optic probe LIBS

Oba, Hironori; Akaoka, Katsuaki; Wakaida, Ikuo; Ouchi, Atsushi*; Mizusako, Fumiki*; Eto, Yoshinori*; Torimaru, Tadahiko*

no journal, , 

no abstracts in English

18 (Records 1-18 displayed on this page)
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