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Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 
1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09
Yamamoto, Masahiro; Sato, Tomonori; Igarashi, Takahiro; Ueno, Fumiyoshi; Soma, Yasutaka
Proceedings of European Corrosion Congress 2017 (EUROCORR 2017) and 20th ICC & Process Safety Congress 2017 (USB Flash Drive), 6 Pages, 2018/09
The authors have studied the differences between outer surface and the crevice-like portion of SUS316L in high pressurized and high temperature water containing dissolved oxygen. We have already introduced that changes in the characteristics of corrosion products along the crevice directions and gap width. It is suggested that the environmental conditions are different with the features of crevice from these results. In this report, we introduce the changes in oxide films with crevice gaps and comparison with the numerical simulation data utilizing of FEM calculation.
Noguchi, Hiroki; Takegami, Hiroaki; Kasahara, Seiji; Tanaka, Nobuyuki; Kamiji, Yu; Iwatsuki, Jin; Aita, Hideki; Kubo, Shinji
Energy Procedia, 131, p.113 - 118, 2017/12
Times Cited Count：22 Percentile：99.79(Energy & Fuels)The IS process is the most deeply investigated thermochemical water-splitting hydrogen production cycle. It is in a process engineering stage in JAEA to use industrial materials for components. Important engineering tasks are verification of integrity of the total process and stability of hydrogen production in harsh environment. A test facility using corrosion-resistant materials was constructed. The hydrogen production ability was 100 L/h. Operation tests of each section were conducted to confirm basic functions of reactors and separators, etc. Then, a trial operation for integration of the sections was successfully conducted to produce hydrogen of about 10 L/h for 8 hours.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 
C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
Soma, Yasutaka; Kato, Chiaki; Ueno, Fumiyoshi
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
Intergranular oxidation (corrosion) occurred within crevice of austenitic low-carbon stainless steel (solution treated, almost no applied stress) after immersion in high temperature water (288C, 8.5 MPa, dissolved oxygen conc. 32 ppm, electrical conductivity: 1.2
0.2
S (measured value at 25C)) for 500 h. The intergranular oxidation occurred at specific position within the crevice that is relatively distant from the crevice mouth with relatively low crevice gap. Both the grain boundary and grain matrix were oxidized. In the oxidized area, Fe and Ni were depleted and Cr was enriched compared to the matrix. Maximum penetration depth of the oxidation was approximately 50 m after 500 h. In order to understand potential-pH condition within the crevice, surface oxide layer was microscopically and thermodynamically investigated. Thermodynamic properties of the surface oxides near the intergranular oxidized area indicated lowered pH of approximately 3.2 to 3.4. In-situ measurement of local solution electrical conductivity was carried out using small electrodes (dia. 800 m) imbedded into the crevice former plate. The solution pH was estimated using theoretically calculated pH vs. electrical conductivity relationship. In the area where the intergranular oxidation occurred, the solution electrical conductivity was nearly 100 times higher than that of bulk water and which indicated lowered pH of approximately 3.5. The above results suggested that, in the high temperature and relatively high purity water, acidification occurs within crevice of stainless steels and such aggressive corrosion condition result in the intergranular oxidation.
Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Uchida, Shunsuke*; Sato, Tomonori; Morishima, Yusuke*; Hirose, Tatsuya*; Miyazawa, Takahiro*; Kakinuma, Nagao*; Sato, Yoshiyuki*; Usui, Naoshi*; Wada, Yoichi*
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.19 - 29, 2005/00
Static and dynamic responses of stainless steel specimens exposed to H
O and O in high temperature water were evaluated by analyzing ECP and FDCI (frequency dependent complex impedance). The oxide films on the specimens were characterized by multilateral surface analyses, e.g., LRS, SIMS, XPS and direct electric resistance measurement. As a result of evaluation, it was confirmed that (1) corrosive condition of BWR normal water chemistry (NWC) was simulated by 100 ppb HO without co-existing O, while that of hydrogen water chemistry (HWC) was simulated by 10 ppb HO, (2) ECP under HWC was as high as that under NWC, while dissolution rate of oxide film under HWC was much lower than that under NWC, (3) combination effects of electric resistance and dissolution rate of oxide caused same level ECP for both NWC and HWC, and (4) distinct weight loss of the specimen exposed to 100 ppb HO was observed.
Nakano, Junichi; Miwa, Yukio; Koya, Toshio; Tsukada, Takashi
Journal of Nuclear Materials, 329-333(Part1), p.643 - 647, 2004/08
Times Cited Count：9 Percentile：48.81(Materials Science, Multidisciplinary)To study effects of minor elements on the irradiation assisted stress corrosion cracking (IASCC), high purity Type 304 and 316 stainless steels (SSs) were fabricated and added minor elements, Si or C. After neutron irradiation to 3.510
n/m
(E
1MeV), the slow strain rate tests (SSRT) for the irradiated specimens was conducted in oxygeneted high purity water at 561 K. Fracture surface of the specimens was examined using the scanning electron microscope (SEM) after the SSRT. Fraction of intergranular stress corrosion cracking (IGSCC) on the fracture surface after the SSRT increased with netron fluence. Suppression of irradiation hardening and increase of peiod to SCC fracture as benefitical effects of the additional elements, Si or Mo, were not observed obviously. In high purity SS added C, fraction of IGSCC was the smallest in the all SSs, although irraidiation hardening level was the largest in the all SSs. Addition of C suppressed the susceptibility to IGSCC.
Nakano, Junichi; Tsukada, Takashi; Tsuji, Hirokazu; Terakado, Shogo; Koya, Toshio; Endo, Shinya
JAERI-Tech 2003-092, 54 Pages, 2004/01
JAERI-Tech-2003-092.pdf:14.05MB
Irradiation assisted stress corrosion cracking (IASCC) is a degradation phenomenon caused by synergy of neutron radiation, aqueous environment and stress on in-core materials, and it is an important issue in accordance with increase of aged light water reactors. Isolating crack initiation stage from crack growth stage is very useful for the evaluation of the IASCC behavior. Hence facility for in-situ observation during slow strain rate test (SSRT) for irradiated material was developed. As performance demonstrations of the facility, tensile test with in-situ observation and SSRT without observation were carried out using unirradiated type 304 stainless steel in 561 K water at 9 MPa. The following were confirmed from the results. (1) Handling, observation and recording of specimen can be operated using manipulators in the hot cell. (2) In-situ observation can be performed in pressurized high temperature water and flat sheet type specimen is suitable for the in-situ observation. (3) Test condition can be kept constantly and data can be obtained automatically for long test period.
Tsukada, Takashi
Zairyo To Kankyo, 52(2), p.66 - 72, 2003/02
Irradiation assisted stress corrosion cracking (IASCC) is a potential failure mode suffered by the core-components of austenitic stainless steels in the aged light-water reactor (LWR), which is the intergranular type cracking caused by synergistic effects of neutron/gamma radiation and chemical environment. Effects of radiation on the materials and high-temperature water are discussed in this paper to understand IASCC phenomenon from a mechanistic viewpoint. It is essential to elucidate the radiation-induced microcompositional and microstructural changes in the alloy for mechanistic and predictive investigations of IASCC. Although grain boundary segregations of alloying and impurity elements are significant factors affecting IASCC, it has been considered that the radiation-induced microstructural and mechanical changes of materials play critical roles in IASCC. For mechanistic understanding of IASCC, further fundamental research works with experimental and theoretical approaches are needed. Efforts directed to the researches at the Japan Atomic Energy Research Institute are also described.
Kusagaya, Kazuyuki*; Sugiyama, Tomoyuki; Nakamura, Takehiko; Uetsuka, Hiroshi
JAERI-Tech 2002-105, 24 Pages, 2003/01
High-temperature and high-pressure influence on the destructive force at the fuel rod failure in reactivity-initiated-accident (RIA) simulating experiment using the NSRR (Nuclear Safety Research Reactor) is estimated, for the purpose of mechanical designing of a new experimental capsule for simulating the temperature and pressure condition of typical commercial BWR. When knowledge on pressure impulse and water hammer, which are the cause of the destructive force, and steam property dependence on temperature and pressure are taken into account, one can qualitatively estimate that the destructive force in the BWR operation condition is smaller than that in the room temperature and atmospheric pressure condition. The water column velocity, which determines the impact by water hammer, is further investigated quantitatively by modeling the experimental system and the water hammer phenomenon. As a result, the maximum velocity of water column in the BWR operation condition is calculated to be only about 10% of that in the room temperature and atmospheric pressure condition.
Tobita, Masahiro*; Itabashi, Yukio
JAERI-Tech 2002-042, 40 Pages, 2002/03
JAERI-Tech-2002-042.pdf:2.09MB
In relation to aging of light water reactors (LWRs), Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded as a significant and urgent issue for reliability of in-core components of LWRs. It is essential for IASCC studies to irradiate test materials under well-controlled of Boiling Water Reactor (BWR) conditions simulating the in-core environment. Therefore, the study for the design of the new water control unit to supply high temperature water into saturated temperature capsules in the Japan Materials Testing Reactor (JMTR) has been carried out. This report summarizes the results of estimation using ORIGEN-2 and QAD-CGGP2 codes of dose equivalent rate on outer surface of the concrete wall of installation room and dose equivalent rate around the ion-exchangers where the highest dose equivalent rate is expected in the unit after the reactor shutdown.
Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*
JAERI-Research 2002-008, 63 Pages, 2002/03
JAERI-Research-2002-008.pdf:7.85MB
no abstracts in English
Tsukada, Takashi; Komori, Yoshihiro; Tsuji, Hirokazu; Nakajima, Hajime; Ito, Haruhiko
Proceedings of International Conference on Water Chemistry in Nuclear Reactor Systems 2002 (CD-ROM), 5 Pages, 2002/00
Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980s and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility.
JAERI-M 85-196, 52 Pages, 1985/12
no abstracts in English
; Kondo, Tatsuo
JAERI-M 83-063, 18 Pages, 1983/04
no abstracts in English
; *
JAERI-M 82-055, 51 Pages, 1982/06
no abstracts in English
Kato, Jun; Taniguchi, Takumi; Osugi, Takeshi; Nakazawa, Osamu; Sone, Tomoyuki; Kuroki, Ryoichiro
no journal, ,
Contaminated water treatment secondary wastes have diversity and have property without processing achievement so far. Tasks for applicability evaluation of these wastes were extracted concerning solidification technologies with applied achievement for radioactive waste processing by FY2017. This report introduces research and development for solving the extracted tasks.
Hirota, Noriaki; Kondo, Keietsu; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*
no journal, ,
Dynamic strain aging (DSA) has been identified in shrouds of boiling water reactors and recirculation system piping of pressurized water reactors in the nuclear field. This phenomenon increases the work hardening rate of the material and causes a reduction in ductility. Rodriguez reported that using stainless steel, this work hardening increases with grain refinement, making DSA more likely to occur. The objective of this study is to evaluate the effect of grain refinement on DSA in a high temperature pressurized water (HTPW) simulating nuclear reactor environment utilizing ultrafine grained SUS304L (UFGS). UFGS was heat treated to adjust the grain size from 0.59 m to 68.6 m, and Hall-Petch relationship for 0.2 % yield stress was arranged. The k values obtained in this study were almost the same as the reference values previously obtained for SUS304L. Regarding the effect of grain size on fracture strain, a comparison of fracture strain between tensile test under air and slow strain rate test (SSRT) under 598 K / 15 MPa at dissolved oxygen 1 ppb showed that the fracture strain was lower than that under air as the grain size became coarser. The micrograph after fracture in a HTPW showed that ductile fracture surfaces were observed for materials with grain sizes less than 28.4 m. However, when the grain size coarsened to 68.6 m, more than half of all fracture surfaces were brittle fractured. For the material with a grain size of 0.59 m under HTPW, many correspondence grain boundaries of {111}/
3 boundaries were observed in the fracture cross-section of the sample. But these distributions were rarely observed when the grain size was coarsened to 68.6 m. Therefore, the suppression of crack propagation by DSA to the fine grains in a HTPW can be attributed to the relaxation of dislocation accumulation by the {111}/3 boundaries.
