High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.

Data processing and visualization of X-ray computed tomography images of a JOYO MK-III fuel assembly

Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji

Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06

https://doi.org/10.1080/00223131.2022.2138603

Irradiation test using foreign reactor, 1; Study of irradiation test with capsule temperature control system (Joint research)

Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.

JAEA-Technology 2022-040, 45 Pages, 2023/03

JAEA-Technology-2022-040.pdf:6.61MB

Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.

Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01

https://doi.org/10.1016/j.jnucmat.2022.154110

Current status of new research reactor at the Monju Site

Mineo, Hideaki

Nihon Genshiryoku Gakkai-Shi ATOMO, 64(11), p.617 - 621, 2022/11

https://doi.org/10.3327/jaesjb.64.11_617

In December 2016 Decisions were made by the Government on the Fast Breeder Prototype Reactor "Monju", which were decommissioning of the reactor and installation of a new research reactor at the Monju site. After the decisions, MEXT started research to list reactor candidates suitable for the site. Among the candidates, medium power reactor type of which thermal output less than 10,000 kW was chosen to utilize neutron beams. Then, from 2020, MEXT launched an entrusted business and adopted JAEA, Kyoto University and University of Fukui as the core institutions of the business to carry out the conceptual design. This paper describes the system to proceed the conceptual design and to examine the utilization management of the new research reactor and also shows their status.

Technical note for the cavitation damage inspection for interior surface of the mercury target vessel, 1; Development of specimen cutting machine for remote handling

Naoe, Takashi; Kinoshita, Hidetaka; Wakui, Takashi; Kogawa, Hiroyuki; Haga, Katsuhiro

JAEA-Technology 2022-018, 43 Pages, 2022/08

JAEA-Technology-2022-018.pdf:7.84MB

In the liquid mercury target system for the pulsed spallation neutron source of Materials and Life science experimental Facility (MLF) at the Japan in the Japan Proton Accelerator Research Complex (J-PARC), cavitation that is generated by the high-energy proton beam-induced pressure waves, resulting severe erosion damage on the interior surface of the mercury target vessel. The erosion damage is increased with increasing the proton beam power, and has the possibility to cause the leakage of mercury by the penetrated damage and/or the fatigue failure originated from erosion pits during operation. To achieve the long term stable operation under high-power proton beam, the mitigation technologies for cavitation erosion consisting of surface modification on the vessel interior surface, helium gas microbubble injection, double-walled beam window structure has been applied. The damage on interior surface of the vessel is never observed during the beam operation. Therefore, after the target operation term ends, we have cut out specimen from the target nose of the target vessel to inspect damaged surface in detail for verification of the cavitation damage mitigation technologies and lifetime estimation. We have developed the techniques of specimen cutting out by remote handling under high-radiation environment. Cutting method was gradually updated based on experience in actual cutting for the used target vessel. In this report, techniques of specimen cutting out for the beam entrance portion of the target vessel in high-radiation environment and overview of the results of specimen cutting from actual target vessels are described.

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

https://doi.org/10.1016/j.anucene.2022.109004

Sorption of Cs and Eu ions onto sedimentary rock in the presence of gamma-irradiated humic acid

Zhao, Q.*; Saito, Takeshi*; Miyakawa, Kazuya; Sasamoto, Hiroshi; Kobayashi, Taishi*; Sasaki, Takayuki*

Journal of Hazardous Materials, 428, p.128211_1 - 128211_10, 2022/04

https://doi.org/10.1016/j.jhazmat.2021.128211

The influence of humic acid and its radiological degradation on the sorption of Cs and Eu by sedimentary rock was investigated to understand the sorption process of metal ions and humic substances. Aldrich humic acid (HA) solution was irradiated with different doses of gamma irradiation using a Co-60 gamma-ray source prior to the contact between the metal ions and the solid sorbent. The HA molecule decomposed to smaller molecules with a lower complexation affinity. Batch sorption experiments were performed to evaluate the effect of gamma-irradiated HA on the sorption of Cs and Eu ions. The addition of non-irradiated HA weakened the sorption of Eu because of the lower sorption of the neutral or negatively charged Eu-HA complexes compared with free Eu ions. The sorption of monovalent Cs ions was barely affected by the presence of HA and its gamma irradiation. The concentration ratio of HA complexed species and non-complexed species in the solid and liquid phases was evaluated by sequential filtration and chemical equilibrium calculations. The ratios supported the minimal contribution of HA to Cs sorption. However, the concentration ratio for Eu in the liquid phase was high, indicating that the complexing ability of HA to Eu was higher than that of HA to Cs ions. Therefore, the sorption of free Eu would predominate with the gamma irradiation dose applied to the HA solution under a radiation field near the HLW package.

Effect of gamma-ray irradiation on corrosion of stainless steel contacted with Zeolite particle

Kato, Chiaki; Yamagishi, Isao; Sato, Tomonori; Yamamoto, Masahiro*

Zairyo To Kankyo, 70(12), p.441 - 447, 2021/12

https://doi.org/10.3323/jcorr.70.441

Zeolite particles have been used in a Cs adsorption vessel for purification of contaminated water in Fukushima Dai-ichi Nuclear Power Station (1F). The used Cs adsorption vessels were kept in storage space on 1F site. The risk of localized corrosion of stainless steel used in the vessel was worried. To evaluate the risk of localized corrosion, using specially designed electrochemical testing apparatus was used under gamma-ray irradiation test. And, real size mock-up test conducted. The results showed the potential change caused by creation of HO by water radiolysis decreased by zeolite particles and the enrichment of chloride ion concentration in the vessel do not propagate during dry up procedure of Cs adsorption vessel. These data indicate the risk of localized corrosion of Cs adsorption vessel may stay at considerably low level.

Evaluation of brittle crack arrest toughness for highly-irradiated reactor pressure vessel steels

Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07