Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Science and Technology, 9 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2243943

The effects of fast neutron irradiation on tensile properties of modified 316 stainless steel (PNC316) claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400 and 735 C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400 and 500C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Swelling behavior of F82H steel irradiated by triple/dual ion beams

Wakai, Eiichi; Kikuchi, Kenji; Yamamoto, Shunya; Aruga, Takeo; Ando, Masami; Tanigawa, Hiroyasu; Taguchi, Tomitsugu; Sawai, Tomotsugu; Oka, Keiichiro*; Onuki, Somei*

Journal of Nuclear Materials, 318, p.267 - 273, 2003/05

https://doi.org/10.1016/S0022-3115(03)00122-3

no abstracts in English

Effect of solute atoms on swelling in Ni alloys and pure Ni under He ion irradiation

Wakai, Eiichi; Ezawa, Tadashi*; Imamura, Junko*; Takenaka, Tsuyoshi*; Tanabe, Tetsuo*; Oshima, Ryuichiro*

Journal of Nuclear Materials, 307-311(Part.1), p.367 - 373, 2002/12

https://doi.org/10.1016/S0022-3115(02)01192-3

no abstracts in English

lrradiation behavior and performance model of nitride fuel

; ;

JNC TN9400 2000-041, 29 Pages, 2000/03

JNC-TN9400-2000-041.pdf:1.18MB

Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.

Irradiation creep of modified 316 and 15Cr-20Ni base austenitic S.S. fuel pins (MFA-1, 2) irradiated in FFTF

; ; Mizuta, Shunji

JNC TN9400 2000-023, 126 Pages, 2000/02

JNC-TN9400-2000-023.pdf:2.94MB

Modified 316 and 15Cr-20Ni base austenitic stainless steels had been developed by Japan Nuclear Cycle Development lnstitute as the candidate materials for Monju and Demonstration fast breeder reactor. Previously, irradiation creep correlation of modified 316 and 15Cr-20Ni had been evaluated using pressurized tubes irradiated in FFTF/MOTA. 0n the other hand, for other austenitic S.S. developed abroad, it was reported that irradiation creep behavior of fuel pin could not be sufficiently described using results of pressurized tube experiments. ln this study, irradiation creep properties of modified 316 and 15Cr-20Ni fuel pins (MFA-I, 2) irradiated in FFTF were evaluated. And irradiation deformation of MFA-1, 2 fuel pins were estimated using the irradiation creep correlation based on MOTA data. The results are summarized as follows : (1)Irradiation creep compliance B calculated from MFA-I, 2 data are 5.6 15.010 [(I0n/m, E>0.1Mev)(MPa)], Which are larger than B based on MOTA data of 2.26.410 and are within the range of B of other austenitic S.S. abroad. (2)Creep-swelling coupling coefficient D derived from MFA-1, 2 data tend to decrease with increasing swelling rate. And the range of D based on MFA-1, 2 data include values calculated from MOTA data of 3.88.210 [(MPa)] and for other austenitic S.S. abroad. (3)As the result that irradiation creep deformation of MFA-1, 2 fuel pins could be appropriately estimated using the irradiation creep correlation derived from MOTA data, it is considered that the creep, correlation based on MOTA data can be applied to estimation of fuel pin deformation.

Irradiation creep equation of the advanced austenitic stainless steels

Mizuta, Shunji; ;

JNC TN9400 99-082, 60 Pages, 1999/10

JNC-TN9400-99-082.pdf:1.52MB

The density measurement of the internal creep specimens irradiated in FFTF/MOTA (Fast Flux Test Facility / Material open Test Assembly) was conducted MMF (Materia1 Monitoring Facility) and accurate separation of swelling strain from total strain leaded in the derivation of the irradiation creep coefficients. Irradiation creep coefficients for PNC 316, 15Cr-20Ni base S.S. and 14Cr-25Ni base S.S. were systematically expressed, while thermal creep coefficients K, under irradiation were separately expressed for above three steels. The results obtained are follows, (1)The effect of stress induced swelling was recognized in the temperature range from 405 to 605C. The swelling in high stress specimens have a tendency to increasing swelling. (2)The irradiation creep coefficients derived from PNC316 and l5Cr-20Ni are similar to that of derived from 20%CW316S.S., CW316Ti and CW15-15Ti which were reported by other authors. (3)The irradiation creep coefficient derived from gas pressurized tube irradiation using FFTF/MOTA expressed appropriately irradiation creep strain from fuel pins using FFTF/MFA-2(15Cr-2ONi base S.S.).

Distributed database system for advanced nuclear materials (Data-Free-Way)

Tsuji, Hirokazu; ; Fujita, Mitsutane*; Kano, Shigeki*; Tachi, Yoshiaki*; Shimura, Kazuki*; Nakajima, Ritsuko*; Iwata, Shuichi*

Advances in Science and Technology, 24, p.417 - 424, 1999/00

no abstracts in English

Development and the results for the control rods in MK-II core of experimental fast reactor JOYO

Miyakawa, Shunichi; ; Soga, Tomonori

PNC TN9410 97-068, 113 Pages, 1997/07

PNC-TN9410-97-068.pdf:3.97MB

Since the first control rod design for the Joyo Mk-II core (about twenty years ago), there have been several challenging improvements; for example, a helium venting mechanism and a flow induced vibration prevention mechanism. Forty-four control rods with these various modifications have been fabricated. To date, thirty-four have been irradiated and the sixteen have been examined, This experience and effort has produced fruitful results: (1)Efficiency and reliability of the diving-bell type Helium venting mechanism (2)Efficiency of the flow induced vibration prevention mechanism (3)Efficiency of the improvement for scram damping mechanism (4)Clarification of absorvber-pellet-cladding-mechanical-interaction (ACMI)phenomena and preventive methods The fourth result listed above has been a subject of investigation for fifteen years in several countries, that is a main phenomena to dominate control rod life time. The results of this investigation of ACMI in absorber elements are summarized below: (a)In five of Joyo Mk-II control rods, cladding cracks were found in fifteen of the elements. These cracks were caused by a acceleration ACMI, due to BC fragments relocation. They occurred over a wide burnup range from 5E+26 Cap./m to 45E+26Cap./m in a nearly typical provability distribution. The cladding cracked because of its low ductility (approximately 1/4 lower than the uniform elongation of usual tensile testing for irradiated 316SS cladding) due to neutron irradiation and the ultra slow ACMI induced strain rate. (b)In this case the crack growth rate is extremely slow and the ACMI induced cracking in absorber elements do not influence either the reactor or plant operations. It is on this basis that a strict limitation to avoid the cladding crack is not necessary. According1y, it is suggested that a realistic design standard should consider the ACMI phenomena and the burnup limit be based on the nominal base calculation for average plastic strain use ...

Behavior of neutron-irradiated USi

Ugajin, Mitsuhiro; Akabori, Mitsuo; Ito, Akinori; Ooka, Norikazu;

Journal of Nuclear Materials, 248, p.204 - 208, 1997/00

https://doi.org/10.1016/S0022-3115(97)00199-2

no abstracts in English

Radiation induced swelling in welded austenitic stainless steel

Sawai, Tomotsugu; Shiba, Kiyoyuki; Fukai, Katsumaro; Hishinuma, Akimichi

6WS-96: 6th Int. Welding Symp. on the Role of Welding Science and Technology in the 21st Century, 2, p.483 - 488, 1996/00

no abstracts in English