Japanese Evaluated Nuclear Data Library version 5; JENDL-5

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.

Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01

https://doi.org/10.1080/00223131.2022.2141903

Comparison of the therapeutic effects of [At]NaAt and [I]NaI in an NIS-expressing thyroid cancer mouse model

Watabe, Tadashi*; Liu, Y.*; Kaneda, Kazuko*; Sato, Tatsuhiko; Shirakami, Yoshifumi*; Oe, Kazuhiro*; Toyoshima, Atsushi*; Shimosegawa, Eku*; Wang, Y.*; Haba, Hiromitsu*; et al.

International Journal of Molecular Sciences (Internet), 23(16), p.9434_1 - 9434_11, 2022/08

https://doi.org/10.3390/ijms23169434

In this study, we compare the therapeutic effect between [At]NaAt and [I]NaI. In vitro analysis of double-stranded DNA break (DSB) and colony formation assay were performed using K1-NIS cells. [At]NaAt induced higher numbers of DSBs and had a reduced colony formation than [I]NaI. In K1-NIS mice, dose-dependent therapeutic effects were observed in both [At]NaAt and [I]NaI. The superior therapeutic effect of [At]NaAt suggests the promising clinical applicability of targeted alpha therapy using [At]NaAt in patients with differentiated thyroid cancer refractory to standard [I]NaI treatment.

Dynamical response of transition-edge sensor microcalorimeters to a pulsed charged-particle beam

Okumura, Takuma*; Azuma, Toshiyuki*; Bennet, D. A.*; Caradonna, P.*; Chiu, I.-H.*; Doriese, W. B.*; Durkin, M. S.*; Fowler, J. W.*; Gard, J. D.*; Hashimoto, Tadashi; et al.

IEEE Transactions on Applied Superconductivity, 31(5), p.2101704_1 - 2101704_4, 2021/08

https://doi.org/10.1109/TASC.2021.3067793

A superconducting transition-edge sensor (TES) microcalorimeter is an ideal X-ray detector for experiments at accelerator facilities because of good energy resolution and high efficiency. To study the performance of the TES detector with a high-intensity pulsed charged-particle beam, we measured X-ray spectra with a pulsed muon beam at the Japan Proton Accelerator Research Complex (J-PARC) in Japan. We found substantial temporal shifts of the X-ray energy correlated with the arrival time of the pulsed muon beam, which was reasonably explained by pulse pileup due to the incidence of energetic particles from the initial pulsed beam.

Effect of coolant water temperature of emergency core cooling system on failure probability of reactor pressure vessel

Lu, K.; Katsuyama, Jinya; Masaki, Koichi; Watanabe, Tadashi*; Li, Y.

Journal of Pressure Vessel Technology, 143(3), p.031704_1 - 031704_8, 2021/06

https://doi.org/10.1115/1.4049048

Comparison of two-phase critical flow models for estimation of leak flow rate through cracks

Watanabe, Tadashi*; Katsuyama, Jinya; Mano, Akihiro

International Journal of Nuclear and Quantum Engineering (Internet), 13(11), p.516 - 519, 2019/10

The estimation of leak flow rates through narrow cracks in structures is of importance for nuclear reactor safety, since the leak flow could be detected before occurrence of loss-of-coolant accidents. The two-phase critical leak flow rates are calculated using the system analysis code, and two representative non-homogeneous critical flow models, Henry-Fauske model and Ransom-Trapp model, are compared. The pressure decrease and vapor generation in the crack, and the leak flow rates are found to be larger for the Henry-Fauske model. It is shown that the leak flow rates are not affected by the structural temperature, but affected largely by the roughness of crack surface.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.

Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12

https://doi.org/10.1007/s11465-018-0487-9

For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.

Analyses of LSTF experiment and PWR plant for 5% cold-leg break loss of coolant accident

Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10

The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.

Loading condition evaluation for structural integrity assessment of RPV due to PTS event based on three-dimensional thermal-hydraulics and structural analyses

Uno, Shumpei; Katsuyama, Jinya; Watanabe, Tadashi*; Li, Y.

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 8 Pages, 2016/07

http://doi.org/10.1115/PVP2016-63443

For structural integrity assessment on reactor pressure vessels (RPVs) of pressurized water reactor during the pressurized thermal shock (PTS) events, thermal history of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analyses simulating PTS events and Jackson-Fewster correlation. Using these values, subsequently, loading conditions for structural integrity assessments of RPVs are evaluated by structural analyses. Nowadays, three-dimensional TH and structural analyses are recognized as precise assessment method for structural integrity of RPVs. In this study, we performed the TH and structural analyses using three-dimensional models of cold-leg, downcomer and RPV in order to evaluate loading conditions during a PTS event more accurately. Distributions of temperature of coolant water and heat transfer coefficient on the surface of RPV were calculated by TH analysis. Loading condition evaluation was then performed by using these values and taking the weld residual stress due to weld-overlay cladding and post-weld heat treatment into consideration. From these analyses, we obtained histories and distributions of loading conditions at the reactor core region of RPV. We discussed the conservativeness of current structural integrity assessment method of RPV prescribed in the current codes through the comparison in the loading conditions due to PTS event.

In-situ measurement of radiation distribution in bottom sediments of irrigation ponds using plastic scintillation fiber

Sanada, Yukihisa; Urabe, Yoshimi; Orita, Tadashi; Takamura, Yoshihide; Torii, Tatsuo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

There are approximately 3,700 irrigation ponds in the Fukushima prefecture. Contamination by radiocesium at the bottom of these ponds has been a concern since the accident at the TEPCO's Fukushima Daiichi Nuclear Power Station (FDNPS). Among them, there are some reservoirs which contain radioactive materials in the bottom sediment. Three years have passed since the accident of the Fukushima Daiichi Nuclear Power Station. In relation to the restoration of agriculture in the disaster-stricken area, there are concerns over the migration of radioactive cesium in the environment. We developed 20 meter-length a plastic scintillation fiber (PSF) for measurement in water. We have conducted tests at dozens of irrigation ponds in Fukushima pref. using PSF to confirm the performance of the devices and to standardize measurement techniques and analytical procedures.

Study on structural integrity assessment of reactor pressure vessel based on three-dimensional thermal-hydraulics and structural analyses

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07

For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.