Inelastic neutron scattering study of magnon excitation by ultrasound injection in yttrium iron garnet

Shamoto, Shinichi; Akatsu, Mitsuhiro*; Chang, L.-J.*; Nemoto, Yuichi*; Ieda, Junichi

Applied Physics Letters, 124(11), p.112402_1 - 112402_5, 2024/03

https://doi.org/10.1063/5.0189768

The magnon excitation by ultrasound injection in YFeO is studied by inelastic neutron scattering. Both longitudinal and transverse ultrasound injections enhanced the inelastic neutron scattering intensity.

Development of an integrated computer code system for analyzing irradiation behaviors of a fast reactor fuel

Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi

Nuclear Technology, 207(8), p.1280 - 1289, 2021/08

https://doi.org/10.1080/00295450.2020.1810977

Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.

Computer code analysis of irradiation performance of axially heterogeneous mixed oxide fuel elements attaining high burnup in a fast reactor

Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*

Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04

https://doi.org/10.1016/j.nucengdes.2019.110448

Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements with high burnup in a fast reactor were conducted. Post-irradiation experiments revealed local concentration of Cs near the interfaces between MOX fuel and blanket columns including the internal blanket of the fuel elements as well as an increase in their cladding diameters. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cs-U-O compound was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

https://doi.org/10.1016/j.nucengdes.2018.03.005

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.

Flexible alpha camera for detecting plutonium contamination

Morishita, Yuki; Yamamoto, Seiichi*; Izaki, Kenji; Kaneko, Junichi*; Nemoto, Norio

Radiation Measurements, 103, p.33 - 38, 2017/08

https://doi.org/10.1016/j.radmeas.2017.04.009

Nuclear fuel materials, such as uranium and plutonium (Pu), are handled at nuclear fuel facilities. There are a contamination source of Pu in tight spaces that cannot be directly measured by a ZnS(Ag) scintillation detector, such as interspace between a glovebox window and a platform, pipe flange, port cover of vinyl bag, and filter interspaces. Therefore, we developed a new imaging detector called a flexible alpha camera that enables to identify the Pu contamination for tight spaces at work sites. The thickness of the flexible alpha camera was only 1/5 of the ZnS(Ag) scintillation detector, and its efficiency for 4-pi direction was 42.7% for 5.5-MeV alpha particles. Minimal detectable activity (MDA) was 0.014 Bq. Four types of PuO sample, removed from a duct, Bag-In/Bag-Out port, glovebox glove, and vinyl sheet, were measured by the flexible alpha camera and the PuO particle was automatically identified. Using the flexible alpha camera, the Pu contamination source can be quickly detected, preventing the spread of contamination.

Verification of the FBR fuel bundle-duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

Uwaba, Tomoyuki; Ito, Masahiro*; Nemoto, Junichi*; Ichikawa, Shoichi; Katsuyama, Kozo

Journal of Nuclear Materials, 452(1-3), p.552 - 556, 2014/09

https://doi.org/10.1016/j.jnucmat.2014.06.016

The BAMBOO code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype FBR and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT)images and local parameters of bundle deformation were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms.

SCC susceptibility of cold-worked stainless steel with minor element additions

Nakano, Junichi; Nemoto, Yoshiyuki; Tsukada, Takashi; Uchimoto, Tetsuya*

Journal of Nuclear Materials, 417(1-3), p.883 - 886, 2011/10

https://doi.org/10.1016/j.jnucmat.2010.12.151

To examine the effects of minor elements on stress corrosion cracking (SCC) susceptibility of low carbon stainless steels with work hardened layer, a high purity type 304 stainless steel was fabricated and minor elements, Si, S, P, C or Ti, were added. Work hardened layer was introduced by shaving on the surface of stainless steels. The specimens were exposed to a boiling 42% MgCl solution for 20 hours and the number and the length of initiated cracks were examined. SCC susceptibility of the specimen with P was the highest and that of the specimen with C was the lowest in all specimens. By magnetic force microscope examination, magnetic phase expected to be martensitic phase was detected near surface. Since corrosion resistance of martensite is lower than that of austenite, the minor elements additions would affect SCC susceptibility through the amount of the transformed martensite, i.e. austenite stability.

In-situ SCC observation of thermally-sensitized and cold-worked type 304 stainless steel irradiated to a neutron fluence of 110n/m

Nakano, Junichi; Nemoto, Yoshiyuki; Miwa, Yukio; Usami, Koji; Tsukada, Takashi; Hide, Koichiro*

Journal of Nuclear Materials, 386-388, p.281 - 285, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.114

Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking on stainless steels were studied by slow strain rate testing in oxygenated high temperature water at 561 K. In-situ observation was carried out during SSRT. Specimens of type 304 stainless steel were subjected to a solution annealing (SA), a thermally sensitization (TS), or a cold working (CW) and irradiated to 1.010 n/m (E 1 MeV) at 323 K in the Japan Material Testing Reactor (JMTR). Crack initiations were observed before the maximum stress would be reached for the CW material in in-situ observation. In fracture surface examination, the TS material exhibited almost intergranular stress corrosion cracking while mixtures of transgranular stress corrosion cracking and ductile dimple fracture were observed for the SA and the CW materials.

PIE technologies for the study of stress corrosion cracking of reactor structural materials

Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio

JAEA-Conf 2006-003, p.253 - 265, 2006/05

https://jopss.jaea.go.jp/pdfdata/JAEA-Conf-2006-003.pdf

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).

JNC-JAERI united research report; A Study on degradation of structural materials under irradiation environment in nuclear reactors (Joint research)

Hoshiya, Taiji*; Ueno, Fumiyoshi; Takaya, Shigeru*; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Aoto, Kazumi*; Tsukada, Takashi; Abe, Yasuhiro*; Nakamura, Yasuo*; et al.

JAERI-Research 2004-016, 53 Pages, 2004/10

JAERI-Research-2004-016.pdf:22.07MB

Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Energy Research Institute (JAERI) have started a JNC-JAERI united research program cooperatively in 2003, which has been aimed for efficient progress and synergistic effect on the research activities of both Institutes in order to lead the facing task of unification between JNC and JAERI. This study has been chosen one of the united research themes, and the purpose of it is to clarify damage mechanism of structural materials under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2003, magnetic flux density distribution (JNC) and micro-corrosion (JAERI) measurement apparatus were newly developed and equipped in Hot Facilities in two Institutes, respectively. These apparatus were designed and produced in consideration of radiation resistance and remote-controlled operation to equip in hot cells. We will start the study on neutron irradiation damage by employing the two apparatus as the next step.