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

Mechanical property evaluation of Zircaloy cladding tube after LOCA-simulated experiment using nanoindentation method (Joint research)

Kakiuchi, Kazuo; Udagawa, Yutaka; Yamauchi, Akihiro*

JAEA-Research 2022-001, 21 Pages, 2022/06

JAEA-Research-2022-001.pdf:1.84MB

The primary cause of cladding embrittlement during loss-of-cool ant accident (LOCA) is the increase in oxygen concentration in the metallic layer and associated microstructural change due to oxidation. In the case of cladding high temperature rupture, inner surface oxidation by the steam ingress and the consequent increase in hydrogen partial pressure result in hydrogen absorption (secondary hydriding) localized in the axial direction at the distance apart from the rupture opening as is well known from preceding studies. In order to understand the effect of cladding microstructural changes on mechanical property of a fuel rod under LOCA conditions in a more precise and quantitative manner, the nanoindentation method has been applied to evaluation of mechanical properties of a cladding specimen after a LOCA simulated test; results for two samples taken from the rupture opening part and secondary hydriding part were compared with each other. The fraction of plastic work during the indentation was evaluated from the load-displacement curve in addition to hardness and Young's modulus. The plastic work fraction at the secondary hydriding part was found to be clearly lower than that at the rupture opening part and rather close to that in the ZrO and -Zr(O) layers, suggesting the significant ductility reduction of the secondary hydriding part despite its relatively low oxygen concentration.

Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Osaka University*

JAEA-Review 2021-046, 77 Pages, 2022/01

JAEA-Review-2021-046.pdf:2.92MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Development of technology to simultaneously measure viscosity and surface tension of molten materials in reactor core" conducted from FY2018 to FY2020. Since the final year of this proposal was FY2020, the results for three fiscal years were summarized. Since (U, Zr)O and boride, molten materials in reactor core, exist at extremely high temperature, chemical reactions between the vessel and these molten materials are unavoidable. Therefore, it is difficult to measure the thermophysical property of these materials. In the present study, droplets are produced by heating and melting the samples levitated by a gas levitation method, then the droplets are collided with a substrate. From the instant behavior of

Nuclide production cross sections of Ni and Zr irradiated with 0.4-, 1.3-, 2.2-, and 3.0-GeV protons

Takeshita, Hayato; Meigo, Shinichiro; Matsuda, Hiroki; Iwamoto, Hiroki; Maekawa, Fujio; Watanabe, Yukinobu*

JPS Conference Proceedings (Internet), 33, p.011045_1 - 011045_6, 2021/03

https://doi.org/10.7566/JPSCP.33.011045

To improve accuracy of nuclear design of accelerator driven nuclear transmutation systems, nuclide production cross sections on Ni and Zr, which were candidate materials to be used in ADS, were measured for GeV energy protons. The measured results were compared with PHITS calculations and JENDL/HE-2007.

Development of HCl-free solid phase extraction combined with ICP-MS/MS for rapid assessment of difficult-to-measure radionuclides, 1; Selective measurement of Zr and Mo in concrete rubble

Do, V. K.; Furuse, Takahiro; Murakami, Erina; Aita, Rena; Ota, Yuki; Sato, Soichi

Journal of Radioanalytical and Nuclear Chemistry, 327(1), p.543 - 553, 2021/01

https://doi.org/10.1007/s10967-020-07503-z

A new HCl-free chromatographic separation procedure has been developed for sequential separation of Zr and Mo from concrete matrices. Accordingly, Zr and Mo could be sensitively and selectively measured by ICP-MS/MS using ammonia reaction gas. The recoveries of greater than 90% for Zr and Mo from concretes could be achieved. The measurement condition was optimized for complete suppression of interferences from Nb and peak tailing from abundant isotopes of Zr and Mo in concrete matrices. The removal of interferences was verified by measurement of radio-contamination-free concretes used as a sample matrix blank. Method detection limits of 1.7 mBq g and 0.2 Bq g were achieved for Zr and Mo, respectively, in the concrete matrices. The interference removal factor for Nb (equivalent to the decontamination factor in radiochemical separation) was of the order of 10, and the abundance sensitivity was of the order of 10, indicating that the developed method is reliable for verifying the presence of ultralow concentrations of Zr and Mo. The present method is suitable for the rapid assessment of Zr and Mo for radioactivity inventory of concrete rubble.

Microscopic analyses on Zr adsorbed IDA chelating resin by PIXE and EXAFS

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nomura, Kazunori; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Hagura, Naoto*; Kubota, Toshio*

Nuclear Instruments and Methods in Physics Research B, 477, p.54 - 59, 2020/08

https://doi.org/10.1016/j.nimb.2019.10.028

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

https://doi.org/10.1016/j.nucengdes.2019.110419

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio

JAEA-Data/Code 2019-018, 22 Pages, 2020/01

JAEA-Data-Code-2019-018.pdf:1.39MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.

Structure change of monoclinic ZrO baddeleyite involving softenings of bulk modulus and atom vibrations

Fukui, Hiroshi*; Fujimoto, Manato*; Akahama, Yuichi*; Sano, Asami; Hattori, Takanori

Acta Crystallographica Section B; Structural Science, Crystal Engineering and Materials (Internet), 75(4), p.742 - 749, 2019/08

https://doi.org/10.1107/S2052520619007923

Monoclinic ZrO baddeleyite exhibits anomalous softenings of bulk modulus and atom vibrations with compression. We have investigated the pressure evolution of the structure by neutron powder diffraction combined with ab-initio calculations. The present results showed that the anomalous pressure response of the bulk modulus is related not to the change in the bonding characters but to the deformation of an oxygen sublattice, especially one of layers made of oxygens in the crystallographic * plane. The layer consists of two parallelograms; one is rotating with little distortion and the other is being distorted with increasing pressure. This deformation of this layer makes one of Zr-O distances long, resulting in the softening of some atom vibrational modes.

High-temperature interaction between zirconium and UO

Shirasu, Noriko; Suzuki, Akihiro*; Nagae, Yuji; Kurata, Masaki

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

High temperature interaction tests between UO and Zr were performed at around 2173 K, to make clear the UO/ -Zr(O) interaction and the mechanism of degradation, for developing the improved models for advanced severe accident analysis codes. A Zr plate was inserted in a UO crucible, and heat treated at 2173 K in stream of Ar. After the heat-treatment, the samples were subjected to surface microanalysis. The middle region of Zr sample shows streak-like structures which are extended towered the top. It is confirmed that the streak-like structures were mainly consist of U from the EDX results, and the structures revealed that the U-rich phase was liquid during the heat-treatment. It seems that the U-rich liquid grew selectively toward the area where the oxygen concentration was low.