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Journal Articles

Austenite-based stainless steel irradiation behavior of the precipitate and void swelling

Inoue, Toshihiko; Sekio, Yoshihiro; Watanabe, Hideo*

Materia, 58(2), P. 92, 2019/02

For the evaluation of irradiated segregation behavior, Austenite-based stainless steel for the fast reactor, during irradiation was evaluated by utilizing TIARA facility (Irradiate temperature: 600 $$^{circ}$$C, Dose: 100 dpa) was observed by analytical electron microscope (JEM-ARM20FC). As a result of observation, the large-size void is observed in irradiation area, and MX segregation (containing Niobium) is not observed. In un-irradiation area the MX segregation is observed. And it is observed conspicuously that Nickel is segregation on the void surface. By the latest high-performance TEM utilization, these phenomenon are able to visualize. It is expected for the clarification of the irradiation damage and mechanism of void swelling, by the analyzing these phenomenon utilization with the latest high-performance TEM utilization.

Journal Articles

Fuel behavior under a severe accident condition

Uetsuka, Hiroshi

Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.156 - 162, 2001/06

no abstracts in English

JAEA Reports

Irradiation tests report of the 35th cycle in "JOYO"

*

JNC-TN9440 2000-008, 79 Pages, 2000/08

JNC-TN9440-2000-008.pdf:2.33MB

This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).

JAEA Reports

Investigation of the properties of high temperature resistance alloys used in the helium gas cooled high temperature reactor

Uwaba, Tomoyuki

JNC-TN9420 2000-005, 28 Pages, 2000/03

JNC-TN9420-2000-005.pdf:0.94MB

In the first phase of the feasibility study, their basic objectives are presentating the feasible image and scenario of development of the FBR cycle system, which is composed of the fast reactor, spent fuel reprocessing and fuel manufacturing facility. In the development of the FBR system in this phase, various ideas of plants are to be studied, which include coolant types such as sodium, heavy metals, gases(CO$$_{2}$$, He), wator, and middle or small size of the reactor, and fuel types (MOX, metal and nitride). In this report, as a part of this study, materials used for the core of the helium gas cooled reactor and their integrity (corrosion, mechanical and irradiation property) under high temperature helium atmosphere were investigated from open literatures.

JAEA Reports

lnvestigation for corrosion behavior of ferritic core materials in C0$$_{2}$$ gas cooled reactor

; ; Mizuta, Shunji

JNC-TN9400 2000-040, 41 Pages, 2000/03

JNC-TN9400-2000-040.pdf:0.85MB

The corrosion behavior of ferritic stainless steels applied to core components under C0$$_{2}$$ gas environment was investigated in order to be helpful to fuel design in C0$$_{2}$$ gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0$$_{2}$$ gas temperature, Si and Cr contents of ferritic steels was determined quantitatively. The following correlations to calculate the metal loss thickness was established. X = 4.4w w = √(k$$times$$t) k = $$alpha$$ $$times$$ exp( - 5.45[Si]) $$times$$ exp( - 1.09[Cr]) $$times$$ exp( - 11253/T) $$alpha$$ = 1.65 $$times$$ 10$$^{8}$$$$sim$$4.40 $$times$$ 10$$^{9}$$ X : metal loss thickness[$$mu$$ml, w : corrosion weight gain [mg/cm$$^{2}$$] k : parabola constant [(mg/cm$$^{2}$$)$$^{2}$$/hr], t : time [hr], $$alpha$$ : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]

JAEA Reports

lnvestigation for corrosion behavior of core materials in lead cooled reactor

Kaito, Takeji

JNC-TN9400 2000-039, 19 Pages, 2000/03

JNC-TN9400-2000-039.pdf:0.66MB

The corrosion behavior of core materials in lead cooled reactor was investigated as the feasibility study for fast breeder reactor. The results are summarized as follows. (1)The corrosion of stainless steels under lead and lithium occurs mainly due to the dissolution of nickel. Consequently ferritic stainless steels have better resistance to corrosion under lead and lithium than austenitic stainless steels, and the corrosion resistance of high nickel steels is worst. (2)The dissolution rate, D(mg/m$$^{2}$$/h), is correlated with lead and lithium temperature, T(K), as log$$_{10}$$ Da = 10.7873 - 6459.3/ T and log$$_{10}$$Df = 7.6185 - 4848.4/T, where D a is the dissolution rate for austenitic steels and D f is for ferritic steels. lt's possible to calculate the corrosion thickness, C($$mu$$m), using the following correlation: C = (D$$times$$t)/$$rho$$$$times$$10$$^{-3}$$, where t is exposure time(hr) and $$rho$$ is density of the core matelial (g/cm$$^{3}$$). (3)The corrosion thickness estimated for austenitic steels using above correlations was extremely larger than ferritic steels, about 6 times at 400$$^{circ}$$C and more than 20 times at above 600$$^{circ}$$C. lt's considered that applicable temperature in lead cooled reactor core is below 400$$^{circ}$$C (about 60$$mu$$m corrosion thickness after 30000 hr) for austenitic steels, and below 500$$^{circ}$$C (about 80 $$mu$$m after 30000 hr) for ferritic steels.

JAEA Reports

Evaluation of charpy impact property in high strength ferritic/martensitic steel (PNC-FMS)

;

JNC-TN9400 2000-035, 164 Pages, 2000/03

JNC-TN9400-2000-035.pdf:3.67MB

High Strength Ferritic/Martensitic Steel (PNC-FMS : 0.12C-11Cr-0,5Mo-2W-0.2V-0.05Nb), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. Ductile brittle transition temperature (DBTT) was tentatively determined in 1992 in material design base standard of PNC-FMS. Howevcr, specimen size effect on impact property and upper shelf energy (USE) have not been evaluated. ln this study, effects of specimen size, thermal aging and neutron irradiation on the charpy impact property of PNC-FMS were evaluated, using together with recently obtained data. The design value of USE and DBTT as fabricated and each correlation of aging and irradiation effects were determined. The results are summarized as follows. (1)lt was found that USE is related to (Bb) as USE=m(Bb)$$^{n}$$, where B is specimen width, b is ligament size and both m and n are constant. For PNC-FMS, n value is equal to 1.4. It's possible to determine n value from USE (J) for full size specimen using the correlation: n=1.38$$times$$10$$^{-3}$$ USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(log$$_{10}$$BKt)+q, where Kt is elastic stress concentration factor and both p and q are constant. For PNC-FMS, the correlation is as follows: DBTT=119(log$$_{10}$$BKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 $$^{circ}$$C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 $$^{circ}$$C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.

JAEA Reports

The survey and evaluation of oxidation for core material of the austenitic stainless steels in carbon dioxide gas-cooled reactor

Mizuta, Shunji; ;

JNC-TN9400 2000-032, 38 Pages, 2000/03

JNC-TN9400-2000-032.pdf:1.2MB

lt is necessary for feasibility study of fast reactor to evaluate the oxidation of the austenitic stainless steels in the case of using for core material in carbon dioxide gas-cooled reactor. The properties for oxidation of austenitic stainless steels in carbon dioxide were surveyed in literatures and the data were selected after evaluation of factors for oxidation in carbon dioxide. The equation of oxidation in carbon dioxide for PE16, 20Cr/25Ni/Nb, 18Cr-8Ni and JNC Cladding materials were proposed. The equation for oxidation of austenitic stainless steels were expressed as upper limit for the equation according to parabolic law. The equation for JNC cladding materials (PNC316, PNC1520, 14Cr-25Ni) was proposed based the oxidation behavior of 18Cr-8Ni which is same oxidation region for weight gain in three-component system of Fe-Cr-Ni, in addition to evaluate of effect for silicon content. The oxidation equation of 20Cr/25Ni/Nb was applied to the high Ni alloy of JNC cladding material. The obtained equation is as follows, X = 4.4W$$times$$1000, W = $$sqrt{(kp・t)}$$, kp = $$alpha$$ exp(-Q/(RT)), X: oxide thickness[$$mu$$m], W : weight gain[g$$times$$cm$$^{-2}$$], kp : parabolic rate constant[g$$^{2}$$$$times$$cm$$^{-4}$$$$times$$ s$$^{-1}$$], t :time[sec] $$alpha$$ : constant[g$$^{2}$$$$times$$cm$$^{-1}$$$$times$$S$$^{-1}$$], Q : activation energy[J・mol$$^{-1}$$], R : gas constant[8.314J $$times$$K$$^{-5}$$ $$times$$mol$$^{-1}$$], T : temperature[K] (1) PE16 : kp = 1.090$$times$$10$$^{-3}$$ exp(-192,500/(RD)), (2) 20Cr/25Ni/Nb : kp = 1.651$$times$$10$$^{-2}$$ exp(-201,300/(RT)) High Ni alloy (JNC), (3)18Cr-8Ni : kp = 1.503$$times$$10$$^{-8}$$ exp(-60,000/(RT)), (4) PNC316, PNC1520 : kp = 1.503$$times$$10$$^{-8}$$ exp(-60,000/(RT))$$times$$0.62$$^{2}$$ 14Cr-25Ni(JNC) The weight gain is (3)$$rangle$$(4)$$rangle$$(2)$$rangle$$(1) in order.

JAEA Reports

Irradiation tests report of the 33rd cycle in "JOYO"

*

JNC-TN9440 2000-002, 157 Pages, 2000/02

JNC-TN9440-2000-002.pdf:5.44MB

This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 33rd cycle, and estimates the 34th cycle irradiation condition. Irradiation tests in the 33rd cycle are as follows: (1)B-type irradiation rig (B9) (a)High burn up performance tests of "MONJU" fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins and large diameter annular pellet fuel pins (b)Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI) (2)C-type irradiation rig (C4F) (a)High burn up performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (3)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (4)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (5)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (6)Core Materials Irradiation Rig (CMIR-5-1) (a)Core materials irradiation tests (7)Structure Materials Irradiation Rigs(SMIR) (a)Material irradiation tests (in collaboration with universities) (b)Surveillance back up tests for "MONJU" (8)Upper core structure Irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burnup driver assembly "PFD516" reached 64,300MWd/t (pin average).

Journal Articles

Corrosion problems on atomic energy plants

Kiuchi, Kiyoshi

Kinzoku, 62(2), p.9 - 15, 1992/02

no abstracts in English

JAEA Reports

None

PNC-TJ8009 91-001, 81 Pages, 1991/06

PNC-TJ8009-91-001.pdf:8.27MB

no abstracts in English

Oral presentation

Progress of R&D and remaining subjects on materials degradation in severe accidents

Nagase, Fumihisa

no journal, , 

JAEA conducts R&D to support the decommissioning at the Fukushima Daiichi NPS and to contribute improvement of the LWR safety in the frame of domestic and international collaborations as well as the own projects. The R&D mostly focuses on the phenomena in BWRs and covers various issues related to materials degradation in severe accidents. In parallel, JAEA has the research activity to establish technical basis for practical use of accident tolerant fuel (ATF) components in existing LWRs. The preliminary computer code analyses showed necessary material data and subjects to design the ATF components.

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