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Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 2; Fuel cladding oxidation

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

Oxidation behaviour of Zr cladding in SFP accident condition was evaluated by using a thermobalance in this work, and the obtained data were applied to construct oxidation model for SFP accident condition. For the validation of the constructed oxidation model, oxidation tests using a long cladding tube 500mm in length were conducted in conditions simulating SFP accidents, such as flow rate of the atmosphere in spent fuel rack, temperature gradient along the axis of cladding, and heating-up history. Thickness of oxide layer formed on the surface of cladding samples was evaluated by cross sectional observation, and compared with calculation results obtained by using the oxidation model. The detail of experimental results and validation of the oxidation model will be discussed.

Journal Articles

Challenge next-generation nuclear system; Development of oxide dispersion strengthened ferritic steel

Otsuka, Satoshi; Kaito, Takeji

Enerugi Rebyu, 39(1), p.44 - 46, 2019/01

For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.

Journal Articles

Oxide dispersion-strengthened/ferrite-martensite steels as core materials for Generation IV nuclear reactors

Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; de Carlan, Y.*; Ribis, J.*; Malaplate, J.*

Structural Materials for Generation IV Nuclear Reactors, p.357 - 414, 2017/00

 Times Cited Count:13 Percentile:3.45

Oxide dispersion strengthened (ODS) steels are the most promising candidate materials for fuel cladding of generation IV nuclear reactors. The progress and current status for development of ODS/FM(ferrite-martensite) steels conducted mainly in Japan and France are overviewed. The chemical compositions of ODS/FM steels are listed. Fabrication routes of cladding tube are mentioned for ferrite-type ODS steels using recrystallized process and martensite-type one using $$alpha$$-$$gamma$$ phase transformation. The optimized process is identical for both countries. Joining process between cladding and end-plug has been also developed by using the pressurized resistance upset welding method. The improvements brought by ODS/FM steels in high-temperature strength and irradiation resistance are verified.

Journal Articles

Study on oxidation behavior of cladding for accident conditions in spent fuel pool

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo*; Tojo, Masayuki*; Goto, Daisuke*

Fushoku Boshoku Kyokai Dai-62-Kai Zairyo To Kankyo Toronkai Koenshu (CD-ROM), p.23 - 24, 2015/11

In order to clarify the air oxidation behavior of the cladding at high temperatures for study on improvement of safety for accident conditions in spent fuel pool, the oxidation tests for both small specimens under constant temperature conditions and long specimens under loss of coolant simulated temperature conditions were carried out, and the knowledge for influence of both temperature gradient and preoxide film on oxidation behavior of the cladding were obtained in this study.

Journal Articles

Analysis of the fracture behavior of hydrided fuel cladding by fracture mechanics

Kuroda, Masatoshi*; Yamanaka, Shinsuke*; Nagase, Fumihisa; Uetsuka, Hiroshi

Nuclear Engineering and Design, 203(2-3), p.185 - 194, 2001/01

 Times Cited Count:13 Percentile:28.3

no abstracts in English

JAEA Reports

Basic evaluation on metal-hydrogen interactions for selecting cladding materials

Ogawa, Hiroaki; Saburi, Tei; Kiuchi, Kiyoshi

JAERI-Research 2000-055, 57 Pages, 2000/11

JAERI-Research-2000-055.pdf:3.73MB

no abstracts in English

JAEA Reports

Evaluation for the transient Burst property of austenitic steel fuel Claddings irradiated as the MONJU type Fuel Assemblies (MFA-1&MFA-2)in FFTF

; ; Sakamoto, Naoki; *; Akasaka, Naoaki;

JNC-TN9400 2000-095, 110 Pages, 2000/07

JNC-TN9400-2000-095.pdf:13.57MB

The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.13$$times$$10$$^{27}$$ n/m$$^{2}$$. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region($$sim$$100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6$$^{circ}$$C. This suggested that the design cladding maximum temperature limit for MONJU (830$$^{circ}$$C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.

Journal Articles

Corrosion behavior of purified Nb alloy and austenitic stainless steel in high temperature and high pressure water

Inohara, Yasuto*; Ioka, Ikuo; Fukaya, Kiyoshi; Kiuchi, Kiyoshi; Kuroda, Yuji*; Miyamoto, Satoshi*

Fushoku Boshoku Kyokai Dai-47-Kai Zairyo To Kankyo Toronkai Koenshu (B-204), p.177 - 180, 2000/00

no abstracts in English

JAEA Reports

None

*; *; *; *; *; *; *

PNC-TJ9009 96-002, 172 Pages, 1995/10

PNC-TJ9009-96-002.pdf:11.22MB

None

Journal Articles

Oxide layer-thickness measurement on fuel cladding for high burnup HBWR fuel rods using eddy current method (Non-destructive examination)

Nakata, Masahito; Amano, Hidetoshi; ; Nishi, Masahiro; Nakamura, Jinichi; Furuta, Teruo; ;

HPR-345, 0, 9 Pages, 1995/00

no abstracts in English

JAEA Reports

None

PNC-TJ1214 94-019, 90 Pages, 1994/06

PNC-TJ1214-94-019.pdf:2.43MB

None

JAEA Reports

Fabrication and evaluation of the tubed functionally gradient material by slurry dipping

Watanabe, Ryuzo*; *

PNC-TJ9601 94-003, 87 Pages, 1994/03

PNC-TJ9601-94-003.pdf:4.58MB

This report is the PNC contract research for fiscal year of 1993 titled "Formation of Ti/SUS/Mo graded layer by slurry dipping." Fuel sheath material for FBR is used under a severe enviroment. The life of conventionally used SUS316 is known to be only two years. The development of long-life core material having high temperature strength, radiation resistance and anti-corrosion property is now essential. To create a super-long-life core materials for FBR it seems promising to employ the concept of functionally gradient material, in which these different materials are configurated with grading : as base material is used SUS316 stainless steel, the inner wall is made of Ti for the radiation resistant and anti-corrosion property with graded intermediate layers towards the base metal and the outer shell is the graded Mo layer for the corrosion resistance against liquid sodium. The shape of the core tube is a long cylindrical tube and its dimensions are 8.5mm in outer diameter, about 2m in length, shell thickness is 0.5mm and the thickness of the gradient layer is about 0.1mm. However, we have not yet acquired sufficient techniques to realize such shape and dimmensions, and the investigation is planned to get basic informations on the processing of the core materials with graded structures. Slurry dipping has been employed for forming a graded layer on curved inner and outer surfaces. And it is indispensable that the graded layers have showed sufficient thermal-stress relief function, as well as good bonding capability. In this year (FY 1993), the formation of the graded inner layer of Ti/SUS316 by slurry dipping was investigated. The formation procedure is briefly mentioned here as follows. Cylindrical green compacts of SUS304 stainless steel powder was prepared by die compaction and CIP as a substrate for slurry dipping. A coarse Ti powder was suspended in ethanol and milled by tumbler ball mill to get a slurry having an appropriate viscosity for dipping. The ...

JAEA Reports

None

Power Reactor and Nuclear Fuel Development Corporation

PNC-TN9360 94-002, 100 Pages, 1994/02

PNC-TN9360-94-002.pdf:4.75MB

no abstracts in English

JAEA Reports

None

Power Reactor and Nuclear Fuel Development Corporation

PNC-TN9360 94-001, 95 Pages, 1994/02

PNC-TN9360-94-001.pdf:4.57MB

no abstracts in English

JAEA Reports

Materials properties data sheet (No.Q 01) ; Internal pressure creep properties data on high strength ferritic/martensitic steel in air and in sodim

; ; *; *; Yoshida, Eiichi;

PNC-TN9450 92-004, 37 Pages, 1992/06

PNC-TN9450-92-004.pdf:0.78MB

High Strength Ferritic/Martensitic Steel is one of the cardidate core materials for largescale FBR because of excellent resistance to swelling. This report are presented about the internal pressure creep of High Strength Ferritic/Martensitic Steel based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; (1) Material: High Strength Ferritic/Martensitic Steel Fuel cladding tube ($$phi$$6.5$$times$$0.47.mm$$^{t}$$) (2) Environment: In Air and In Sodium (3) Test temperature: 600 and 650$$^{circ}$$C (4) Hoop stress: 9.48$$sim$$32.43 kgf/㎜$$^{2}$$ (5) Number of data: 13 points

JAEA Reports

None

*; *; *; *; *; *; *

PNC-TJ9009 91-004, 149 Pages, 1991/08

PNC-TJ9009-91-004.pdf:24.83MB

None

JAEA Reports

None

Tsutagi, Koichi; ; Tobita, Noriyuki; ; *; *

PNC-TN8410 91-174, 40 Pages, 1991/02

PNC-TN8410-91-174.pdf:5.06MB

None

JAEA Reports

Dissolution studies of spent nuclear fuels

JAERI-M 91-010, 187 Pages, 1991/02

JAERI-M-91-010.pdf:9.73MB

no abstracts in English

Journal Articles

Distributions of radionuclides onand in spent nuclear fuel claddings of pressurized water reactor

Hirabayashi, Takakuni; Sato, Tadashi; Sagawa, Chiaki; ; Saeki, Masakatsu; Adachi, Takeo

Journal of Nuclear Materials, 174, p.45 - 52, 1990/00

 Times Cited Count:16 Percentile:17.56

no abstracts in English

JAEA Reports

Traial Irradiation in JMTR,Part IV; Zirconium and other Materials

; ;

JAERI-M 5648, 110 Pages, 1974/03

JAERI-M-5648.pdf:5.05MB

no abstracts in English

63 (Records 1-20 displayed on this page)